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Sample records for 95zr 103ru 106ru

  1. 103Ru for tumor scanning, 2

    The mechanism of 103Ru-uptake in tumors was investigated through the incubation of rat ascites hepatoma cells (AH-130) in vitro with various concentrations of Ru-chloride containing 103Ru-chloride as a tracer. Quantitative analysis of Ru binding to the cells indicated that ascites hepatoma cells contained high- and low-affinity binding sites for Ru. When ascites hepatoma cells were incubated with Ru after incubation with a low concentration of papain, most of the Ru was not bound to the cells but was found in the medium containing solubilized glycoproteins. However Ru bound mainly to washed cells after the incubation with papain. About 65% of the Ru bound to ascites hepatoma cells was liberated by the papain treatment, and about 45% of the liberated Ru was precipitated by cetyltrimethylammonium bromide, indicating that Ru bound tightly to glycopeptides. These results suggest that the tumor affinity of 103Ru is related to specific binding to glycopeptides on the tumor cell surface. (author)

  2. Absolute standardization of 106Ru by anti-coincidence method

    The system of absolute standardization activity of radionuclide by anti-coincidence counting and live-time techniques was implemented at LNMRI in 2008 to reduce the impacts of some influence factors in the determination of the activity with coincidence counting technique used for decades in the lab, for example, the measurement time. With the anti-coincidence system, the variety of radionuclides that can be calibrated by LNMRI was increased, in relation to the type of decay. The objective of this work is the standardization of 106Ru by the method of counting anti-coincidence and estimate its measurement uncertainties. (author)

  3. β-Ray brachytherapy with 106Ru plaques for retinoblastoma

    Purpose: A retrospective analysis of 134 patients who received 106Ru brachytherapy for retinoblastomas (175 tumors in 140 eyes). Treatment and follow-up were analyzed with special emphasis on tumor control organ, preservation, and late complications. Results: Treated tumors had a mean height and diameter of 3.7 ± 1.4 mm and 5.0 ± 2.8 disk diameters, respectively. The radiation dose values were recalculated according to the calibration standard recently introduced by the National Institute of Standards and Technology. The recalculation revealed a mean applied dose of 419 Gy at the sclera (SD, 207 Gy) and 138 Gy (SD, 67 Gy) at the tumor apex. The 5-year tumor control rate was 94.4%. Tumor recurrence was more frequent in eyes with vitreous tumor cell seeding or fish-flesh regression. The estimated 5-year eye preservation rate was 86.5%. Previous treatment by brachytherapy or external beam radiotherapy, as well as a large tumor diameter, were significant factors for enucleation. The radiotherapy-induced complications after 5 years of follow-up were retinopathy (22%), optic neuropathy (21%), and cataract (17%). These complications were significantly more frequent after prior brachytherapy or external beam radiotherapy. Conclusion: Brachytherapy using 106Ru plaques is a highly efficient therapy with excellent local tumor control and an acceptable incidence of side effects

  4. Transfer from soil to plants of 106Ru as nitrosyl and as chloride

    The transfer of 106Ru in a soil-plant ecosystem was investigated with respect to two chemical forms in compact soil samples under greenhouse conditions with surface and deep-layer contamination. Considerable differences in the uptake of 106Ru were observed between 106RuCl3 and 106Ru-nitrosyl during the first 5-8 wk after the contamination of the soil. The translocation of 106Ru in the soil showed an inhomogeneous distribution of the radioruthenium, with a great part of the total activity remaining in the upper soil layer between 0 and 5 cm even 10 mo after contamination of the soil surface. During the whole experiment, reemission of 106Ru into the air was investigated by using special air collectors under different temperature and light conditions. Although a continuous checking out for a time of about 8 mo, no measurable concentrations of 106Ru could be out for a time of about 8 mo, no measurable concentrations of 106Ru could be found in examined air filters

  5. Clinical quality assurance for 106Ru ophthalmic applicators

    Background and purpose: Episcleral brachytherapy using 106Ru/106Rh ophthalmic applicators is a proven method of therapy of uveal melanomas sparing the globe and in many cases sparing the vision. In the year 2001, an internal clinical quality assurance procedure revealed that part of the ophthalmic applicators leaked and that the calibration was erroneous. Consequently, the producer modernized its production procedures and, in May 2002, introduced a dose rate calibration that is traceable to the NIST standard. This NIST calibration confirmed that the previous calibration had been incorrect. In order to study the effects of the producer's new internal quality assurance procedures on the ophthalmic applicators, applicators of this new generation were submitted to a newly improved internal clinical acceptance test. Patients and methods: The internal clinical acceptance test consists of a leakage test and a dosimetric test of the ophthalmic applicators. The leakage test simulates contact of the ophthalmic applicators with chloride containing body fluid. The dosimetric tests measure depth dose curves and dose rate with a plastic scintillator dosimetric system and compare them with the indications in the producer's certificate. Furthermore, the depth dose profile of the most frequently used applicator (type CCB) was compared with published data. Results: The internal clinical leakage test showed that all of the tested ophthalmic applicators belonging to the new generation (n=17) were tight and not contaminated. The dosimetric acceptance tests applied to seven different types of applicators revealed that the relative depth dose profiles in the therapeutically relevant range (up to a depth of ≤7 mm) deviate from the producer's indications only by -2.7 to +3.2%. The acceptance test of the dose rate values of the ophthalmic applicators at a distance of 2 mm from the surface of the applicators resulted in a coefficient of variation of 1.7% (n=17). In the evaluation of the

  6. Inter-taxa differences in root uptake of 103/106Ru by plants

    Ruthenium-106 is of potential radioecological importance but soil-to-plant Transfer Factors for it are available only for few plant species. A Residual Maximum Likelihood (REML) procedure was used to construct a database of relative 103/106Ru concentrations in 114 species of flowering plants including 106 species from experiments and 12 species from the literature (with 4 species in both). An Analysis of Variance (ANOVA), coded using a recent phylogeny for flowering plants, was used to identify a significant phylogenetic effect on relative mean 103/106Ru concentrations in flowering plants. There were differences of 2465-fold in the concentration to which plant species took up 103/106Ru. Thirty-nine percent of the variance in inter-species differences could be ascribed to the taxonomic level of Order or above. Plants in the Orders Geraniales and Asterales had notably high uptake of 103/106Ru compared to other plant groups. Plants on the Commelinoid monocot clades, and especially the Poaceae, had notably low uptake of 103/106Ru. These data demonstrate that plant species are not independent units for 103/106Ru concentrations but are linked through phylogeny. It is concluded that models of soil-to-plant transfer of 103/106Ru should assume that; neither soil variables alone affect transfer nor plant species are independent units, and taking account of plant phylogeny might aid predictions of soil-to-plant transfer of 103/106Ru, especially for species for which Transfer Factors are not available

  7. Study on the characteristics of indigenously developed 106Ru/106Rh source

    A pilot study was initiated to investigate the characteristics of 106Ru/106Rh source for the purpose of inducting it in Quality Assurance programme of personnel monitoring by Bhabha Atomic Research Centre. Towards this goal, a prototype 106Ru/106Rh source was fabricated indigenously. Different parameters such as activity, gamma ray spectrum, beam uniformity, percentage depth dose, output of the source, contribution of gamma and beta components of the source were studied. The detectors used in the study are CaSO4:Dy Teflon disc, pocket dosimeter, HPGe detector and radiochromic films. - Highlights: • Indigenous development of 106Ru/106Rh source for QA in personnel monitoring. • Source parameters such as activity of the source, gamma spectrum were studied. • Dosimetric parameters such as beam uniformity, BSF, gamma and beta contribution and PDD were studied. • TLDs, EBT radiochromic film, HPGe and electronic pocket dosimeter were used in the study

  8. Numerical calculation of relative dose rates from spherical 106Ru beta sources used in ophthalmic brachytherapy

    Eduardo de Paiva

    2015-01-01

    Full Text Available Concave beta sources of 106Ru/106Rh are used in radiotherapy to treat ophthalmic tumors. However, a problem that arises is the difficult determination of absorbed dose distributions around such sources mainly because of the small range of the electrons and the steep dose gradients. In this sense, numerical methods have been developed to calculate the dose distributions around the beta applicators. In this work a simple code in Fortran language is developed to estimate the dose rates along the central axis of 106Ru/106Rh curved plaques by numerical integration of the beta point source function and results are compared with other calculated data.

  9. Numerical calculation of relative dose rates from spherical 106Ru beta sources used in ophthalmic brachytherapy

    de Paiva, Eduardo

    Concave beta sources of 106Ru/106Rh are used in radiotherapy to treat ophthalmic tumors. However, a problem that arises is the difficult determination of absorbed dose distributions around such sources mainly because of the small range of the electrons and the steep dose gradients. In this sense, numerical methods have been developed to calculate the dose distributions around the beta applicators. In this work a simple code in Fortran language is developed to estimate the dose rates along the central axis of 106Ru/106Rh curved plaques by numerical integration of the beta point source function and results are compared with other calculated data.

  10. Shortcomings of the industrial quality assurance of 106Ru ophthalmic plaques

    Background: Beta emitting 106Ru applicators manufactured by Bebig GmbH (Berlin, Germany) are widely used to treat intraocular tumors. The applicators are fixed to the bulbus and removed after several days. The following therapy relevant defects have been detected by an internal clinical acceptance test: risk of leakage and inconsistent dose-rate specifications by the manufacturer. In the meantime, components of the internal clinical acceptance test have been adopted successfully by the manufacturer of the 106Ru ophthalmic plaques. Material and Method: 106Ru ophthalmic plaques were tested with the following internal clinical acceptance tests: visual inspection, surface contamination, leakage, and dose-rate verification. The surface contamination test consists of a wet wipe test at moderate pressure. For the leakage test of the 106Ru ophthalmic plaques a clinically relevant scenario was developed in which the contact of the applicator with human tissue is simulated. In the course of it the applicator is inserted into Ringer's solution for several days. The certified energy dose-rate statements of the manufacturer are examined with a 1 mm3 plastic scintillator for consistency. (orig.)

  11. Internal clinical acceptance test of the dose rate of 106Ru/106Rh ophthalmic applicators

    For the last forty years or so, episcleral brachytherapy using 106Ru/106Rh ophthalmic applicators has been a proven method of therapy for uveal melanomas, sparing the globe and in many cases conserving the vision. In episcleral brachytherapy, a radioactive 106Ru/106Rh ophthalmic applicator (BEBIG Co., Berlin, Germany) is temporarily fixed on the surface of the bulbus oculi, whereby the intraocular tumour gets irradiated protractedly through the sclera. 106Ru/106Rh ophthalmic applicators are primarily beta sources, i.e. they generate a local dose escalation both in the vessels supplying the tumour and in the tumour itself, while simultaneously sparing the risk structures. In its certificates, BEBIG, the manufacturer of the product, indicates a dose rate for the 106Ru/106Rh ophthalmic applicators at a dose specification reference point which ensures traceability to the NIST standard (12/2001). The dose specification reference point is situated at a distance of 2 mm from the middle of the inner (concave) surface of the applicators, and the dose rate was measured with a scintillation detector (diameter 1 mm, height 0.5 mm). The manufacturer indicates for this dose rate at the dose specification reference point a relative measurement uncertainty of ±20% within the 95% confidence interval. Since the introduction of the NIST calibration, the quality of the calibration passed on by BEBIG to the user has been examined for n=45 ophthalmic applicators

  12. Separation and determination of 103Ru in samples of fission 99Mo

    In Argentina 99Mo is produced in the RA-3 reactor at the Ezeiza Atomic Center (CAE), by irradiation of miniplates of Al/U (90% 235U) alloy. The 99Mo separation is carried out at the Fission Radioisotopes Production Plant. Quality control is important to assure the quality of molybdenum that is produced in CAE. A new method to purify and on line quantify 103Ru as an impurity present in 99Mo samples was developed. This procedure is based in the RuO4 volatilization and its dissolution in NaOH 6M. This is necessary due to the fact that 103Ru cannot be detected in presence of high activities of 99Mo without previous separation. This method allows a quantitative, specific, efficient, fast and reproducible separation of 103Ru from 99Mo. (author)

  13. Organ distribution of biogenic amine derivatives of 103Ru labelled ruthenocenyl - radiopharmaceuticals for adrenal and ovar

    The organ distribution of 103Ru labelled ruthenocenyl derivatives of tyramine, histamine, benzylamine, phenylethylamine and homoveratrylamine were measured in rats. The derivatives of tyramine, histamine and benzylamine showed a high affinity for the adrenal and ovar. Adrenal/muscle ratios up to 2000/1 were gained but only if the dose was administered i. v. and was below 0,1 μmol/kg. The ruthenocenyl derivatives of tyramine labelled with 103Ru in the ruthenocene moiety or with 14C in the tyramine moiety showed a parallel distribution pattern but completely different from the distribution of 103RuCl3. This indicates that the tyramine derivatives are not destroyed in the body yielding Ru-ions. The advantages of the ruthenocenyl derivatives in comparison with the known amphetamine derivatives labelled with radioactive iodine are discussed. (orig.)

  14. Tissue distributions of 97Ru and 103Ru in subcutaneous tumor of rodents

    Tanabe,Masatada

    1975-12-01

    Full Text Available Mice bearing Ehrlich tumor were administered 97Ru-chloride or 103Ru-chloride intravenously. Examinations of various tissues indicated similar distributions by the two radionuclides. The levels were higher in the lung, liver and kidney than in the tumor tissue. Rats bearing AH-130 tumor were administered 103Ru-chloride intravenously. The 103Ru distribution in rats was highest in the spleen, followed by the liver and kidney; however, the radioactive distribution in the tumor tissue exceeded the muscle level by about 5-fold. Tumors were delineated in rats by scintigraphy. The findings indicate that ruthenium radionuclides may be a useful clinical agent in the delineation of some types of tumors. Ruthenium-97 would be favored in possible clinical usage due to its shorter physical half-life and lower levels of gamma energy.

  15. Dynamics of transfer and distribution of 95Zr in the broadbean-soil ecosystem

    The transfer and distribution of 95Zr in a simulated broadbean-soil system was studied by using isotope-tracer techniques. The results showed that the 95Zr was mainly concentrated in the haulm, pod and root, and the activity concentration of 95Zr in these tissues reached the maximum in the initial stage then decreased continuously. The activity concentration of 95Zr in edible part-bean was relatively lower, which was just near to the detection limit. The 95Zr in soil was mainly (97%) deposited in surface layer soil (0-6 cm), indicating that the 95Zr absorbed by surface soil could not be moved downwards easily because of the strong adsorption. The dynamics of 95Zr concentrations in broadbean and soil were also confirmed by application of nonlinear regression method

  16. Internal Clinical Acceptance Test of the Dose Rate of 106Ru/106Rh Ophthalmic Applicators

    Episcleral brachytherapy using 106Ru/106Rh ophthalmic applicators is a proven method of therapy for uveal melanomas, sparing the globe and in many cases, conserving vision. In its certificates, Bebig, the manufacturer of the product, indicates a dose rate for the 106Ru/106Rh ophthalmic applicators which ensures traceability to the NIST standard (12/2001). Since the introduction of the NIST calibration, the quality of the calibration provided by Bebig to the clinical user has been examined for 45 ophthalmic applicators with a plastic scintillator measurement system. Of these, 20 ophthalmic applicators had a dose rate at the dose specification reference point that exceeded the dose rate stated in the manufacturer's certificate by up to 23%. (author)

  17. Retention and excretion of 95Zr-95Nb in humans.

    Thind, K S

    1995-12-01

    This note describes the retention and excretion of 95Zr-95Nb in humans based on a recent CANDU experience and a literature survey of reported cases. Two data bases, QUEST and INIS were used for the survey. Three reported cases were discovered: two for occupational exposures and one for public exposure from nuclear weapons fallout. Human lung retention from these three cases, plus whole body retention and some limited fecal excretion data from a recently occurred exposure at a CANDU station, were reviewed and tested against predictions based on ICRP Publication 30 model. Based on the fits of this model to the reported data it seems that the three occupational exposures exhibit class Y behavior while the public exposure exhibits class W behavior. For only one case is the chemical compound known with certainty: ZrO2. Zirconium oxides are currently classified as class W in ICRP Publications 30 and 54. This work confirms a suggestion that oxides of zirconium be classified as class Y and should be taken into account by the ICRP in its future publications. PMID:7493813

  18. Burn-up cross sections of 51Cr, 59Fe, 65Zn, 86Rb, 103Ru

    Targets of Cr, Fe, Zn, Rb, and Ru were irradiated in the hydraulic tube of the Oak Ridge HFIR reactor at a neutron flux of 2.6 x 1015 n/cm2sec for 1 day and 20 days. The reactor burn-up cross sections (in barns) of the radioactive product nuclides are: 51Cr, 59Fe, 65Zn, 60 +- 30; 86Rb, 103Ru, <20

  19. On the actual state of industrial quality assurance procedures with regard to 106Ru ophthalmic plaques

    In the year 2002, Bebig updated, among other things, the ASMW (GDR) calibration of the dose rate of the 106Ru ophthalmic plaques from the years 1987-1989 by a calibration of the NIST (USA). The current NIST calibration, together with the new equipment for the measurement of the depth dose curves, led to the consequence that the new NIST 2001 dose rate values show, in the mean, a deviation of 0.75 times (plaque type CCC) up to 2.06 times (plaque types CCX, CCY, and CCZ) compared to the dose rate values that had been indicated so far in Bebig's certificate, based on the ASMW 1987 calibration. For the 95% confidence interval, Bebig estimated the measurement uncertainty to be ± 25%. If one takes into consideration the minimal and maximal values in such 95% confidence intervals, it follows that the new NIST 2001 dose rate values deviate between 0.56 times (plaque type CCC) and 2.58 times (plaque types CCX, CCY, and CCZ) from the Bebig certificate (ASMW calibration 1987). As regards leakage, no objections arose in the case of the 106Ru ophthalmic plaques produced according to the new quality standards. (orig.)

  20. Assessment, modelization and analysis of 106 Ru experimental transfers through a freshwater trophic system

    Experiments are carried out in order to study 106 RU transfers through a freshwater ecosystem including 2 abiotic compartments (water and sediment) and 3 trophic levels (10 species). Experimental results are expressed mathematically so as they can be included into a global model which is then tested in two different situations. The comparison of the available data concerning the in situ measured concentrations to the corresponding calculated ones validates the whole procedure. Analysis of the so validated results lightens ruthenium distribution process in the environment. The rare detection of this radionuclide in organisms living in areas contaminated by known meaningful releases can be explained by a relativity high detection limit and by a slight role of the sediment as a secondary contamination source. (author). 78 figs., 18 tabs

  1. Responses of different dosemeters in beta dosimetry of 106Ru/106Rh ophthalmic applicators

    This work presents the TL response of three kinds of dosimeters from different manufacturing characteristics under irradiation of 106 Ru / 106 Rh sealed sources used in ophthalmic brachytherapy. They are: Ca SO4:Dy + teflon (D- Ca SO4:Dy -0,4), LiF:Mg, Ti (TLD-100) and Ca SO4:Dy (TLD-900). Some of reports accepted by scientific community (NCS report 14 e ICRU report 72) as reference in the quality control of beta applicators dosimetry recommend that the absorbed dose standard uncertainties can be kept below 20%. The TLD Ca SO4:Dy + teflon presented proper sensibility and high precision comparing with the others. Considering the similar dimensions of ophthalmic tumors and aside critical structures it is relevant to reduce undesirable effects due to the irradiation of these structures. Therefore, the quality control in the beta dosimetry using this kind of source is a constant challenge. (author)

  2. Accurate estimation of dose distributions inside an eye irradiated with 106Ru plaques

    Background: Irradiation of intraocular tumors requires dedicated techniques, such as brachytherapy with 106Ru plaques. The currently available treatment planning system relies on the assumption that the eye is a homogeneous water sphere and on simplified radiation transport physics. However, accurate dose distributions and their assessment demand better models for both the eye and the physics. Methods: The Monte Carlo code PENELOPE, conveniently adapted to simulate the beta decay of 106Ru over 106Rh into 106Pd, was used to simulate radiation transport based on a computerized tomography scan of a patient's eye. A detailed geometrical description of two plaques (models CCA and CCB) from the manufacturer BEBIG was embedded in the computerized tomography scan. Results: The simulations were firstly validated by comparison with experimental results in a water phantom. Dose maps were computed for three plaque locations on the eyeball. From these maps, isodose curves and cumulative dose-volume histograms in the eye and for the structures at risk were assessed. For example, it was observed that a 4-mm anterior displacement with respect to a posterior placement of a CCA plaque for treating a posterior tumor would reduce from 40 to 0% the volume of the optic disc receiving more than 80 Gy. Such a small difference in anatomical position leads to a change in the dose that is crucial for side effects, especially with respect to visual acuity. The radiation oncologist has to bring these large changes in absorbed dose in the structures at risk to the attention of the surgeon, especially when the plaque has to be positioned close to relevant tissues. Conclusion: The detailed geometry of an eye plaque in computerized and segmented tomography of a realistic patient phantom was simulated accurately. Dose-volume histograms for relevant anatomical structures of the eye and the orbit were obtained with unprecedented accuracy. This represents an important step toward an optimized

  3. Monte Carlo calculation of dose to water of a 106Ru COB-type ophthalmic plaque

    The concave eye applicators with 106Ru/106Rh or 90Sr/90Y beta-ray sources are worldwide used in brachytherapy for treating intraocular tumors. It raises the need to know the exact dose delivered by beta radiation to tumors but measurement of the dose to water (or tissue) is very difficult due to short range of electrons. The Monte Carlo technique provides a powerful tool for calculation of the dose and dose distributions which helps to predict and determine the doses from different shapes of various types of eye applicators more accurately. The Monte Carlo code MCNPX has been used to calculate dose distributions from a COB-type 106Ru/106Rh ophthalmic applicator manufactured by Eckert and Ziegler BEBIG GmbH. This type of a concave eye applicator has a cut-out whose purpose is to protect the eye nerve which makes the dose distribution more complicated. Several calculations have been performed including depth dose along the applicator central axis and various dose distributions. The depth dose along the applicator central axis and the dose distribution on a spherical surface 1 mm above the plaque inner surface have been compared with measurement data provided by the manufacturer. For distances from 0.5 to 4 mm above the surface, the agreement was within 2.5% and from 5 mm the difference increased from 6% up to 25% at 10 mm whereas the uncertainty on manufacturer data is 20% (2s). It is assumed that the difference is caused by nonuniformly distributed radioactivity over the applicator radioactive layer

  4. Process optimization for effective column separation of 106Ru from aqueous waste associated with spent reprocessing solvent in storage tanks

    The present work deals with another waste stream resulting from reprocessing operations, viz. the aqueous solution present in substantial quantities as the bottom layer in tanks storing spent TBP-dodecane solvent. The effective separation of 106Ru from aqueous waste streams generated during reprocessing of spent nuclear fuel is difficult because of its complex aqueous chemistry

  5. Transfer of Chernobyl-derived 134Cs, 137Cs, 131I and 103Ru from flowers to honey and pollen

    The activity concentrations of 137Cs, 134Cs, 131I and 103Ru were determined separately in honey and pollen samples collected from a single bee colony during several months after the deposition of Chernobyl fallout. The source of each honey and pollen sample was determined by pollen analysis. Although the activity concentrations in honey and pollen varied with time, the concentrations of 137Cs and 134Cs were, in general, higher in pollen than in honey. For 103Ru and 131I, these differences were comparatively small. The mean 131I/137Cs and 103Ru/137Cs ratios were about one order of magnitude higher in honey than in pollen. The mean 131I/103Ru ratio was about the same for honey and pollen. This observation, in the light of the corresponding nuclide ratios found in the deposition, suggests that 137Cs, 134Cs, 131I and 103Ru were taken up by the plant leaves and transported to nectar and pollen. The higher activity concentrations of 137Cs and 134Cs in pollen, relative to honey, indicate that these radionuclides behave analogously to potassium, which is also found in higher quantities in pollen. (author)

  6. 129I, 60Co, and 106Ru measurements on water samples from the Hanford project environs

    Groundwater flow and contamination patterns beneath the Hanford project reservation have been studied since the early days of the project. The measurement of radioactive materials at concentrations much below those required for radiation protection are useful for tracing groundwater movement and detection of potential contamination problems before they are apt to occur. Groundwater samples from a number of wells on or near the Hanford reservation have been analyzed for 129I by neutron activation analysis and for gamma radioactivity by low-level coincidence gamma-ray spectrometry. The major radionuclides in addition to natural radioactivity detected in the underground waters by gamma-ray spectrometry were 106Ru and 60Co. Local river and rain water samples were also analyzed for 129I and long-lived radionuclides. Special sample collection methods were developed to prevent contamination of the water samples during collection. Anions travel farther than cations in underground water systems since soils are primarily cation exchangers and retain the cations. Anion exchange techniques were used in the field and the laboratory to recover the desired radionuclides. Sample sizes ranged up to several thousand liters. This paper discusses the sample collection methods,analysis methods, and results obtained. The methods used were found to provide high sensitivity for groundwater studies. (auth)

  7. Studies on the chemical behaviour of 106Ru in sea water and its uptake by marine organisms

    In the previous paper, the author represented that the concentration factor of 106Ru was relatively high for marine algae in comparison with the other organisms such as fish and mollusca. The concentration seemed to be attributable to the effect of adsorption on the surface of the marine algae. And it is well known that the surface substance is quite different according to the species of marine algae. Then uptake behaviour of the ionic species obtained from 106RuNO.Cl sub(x) were investigated in relation to the species of marine algae and the surface substances extracted from them. The cationic complex species represented the highest concentration factors for the two species except red algae. The order of the concentration factors among these was as follows: brown algae (Hizikia fusiforme) > green algae (Ulva pertusa) > red algae (Porphyra tenera). (auth.)

  8. Ferrocene, ruthenocene or rhodocene analogues of Haloperidol. Synthesis and organ distribution after labelling with 103Ru or 103mRh

    Ferrocene-Haloperidol was synthesized by N-alkylation of 4-(4'-chlorophenyl)-4-hydroxypiperidine with 1-ferrocenyl-4-chlor-butan-1-on. By heating the ferrocene-haloperidol with 103RuCl3 the 103Ru-labelled ruthenocene-haloperidol was obtained. This compound showed a high affinity for lung but not for brain in rats and mice. The decay of the 103Ru labelled compound results in the formation of the 103mRh labelled rhodocene-haloperidol, which is rapidly oxidized by air to the corresponding rhodocinium-haloperidol. This compound can be separated by extraction and TLC. (author)

  9. Assessment of ocular beta radiation dose distribution due to 106Ru/106Rh brachytherapy applicators using MCNPX Monte Carlo code

    Nilseia Aparecida Barbosa

    2014-08-01

    Full Text Available Purpose: Melanoma at the choroid region is the most common primary cancer that affects the eye in adult patients. Concave ophthalmic applicators with 106Ru/106Rh beta sources are the more used for treatment of these eye lesions, mainly lesions with small and medium dimensions. The available treatment planning system for 106Ru applicators is based on dose distributions on a homogeneous water sphere eye model, resulting in a lack of data in the literature of dose distributions in the eye radiosensitive structures, information that may be crucial to improve the treatment planning process, aiming the maintenance of visual acuity. Methods: The Monte Carlo code MCNPX was used to calculate the dose distribution in a complete mathematical model of the human eye containing a choroid melanoma; considering the eye actual dimensions and its various component structures, due to an ophthalmic brachytherapy treatment, using 106Ru/106Rh beta-ray sources. Two possibilities were analyzed; a simple water eye and a heterogeneous eye considering all its structures. Two concave applicators, CCA and CCB manufactured by BEBIG and a complete mathematical model of the human eye were modeled using the MCNPX code. Results and Conclusion: For both eye models, namely water model and heterogeneous model, mean dose values simulated for the same eye regions are, in general, very similar, excepting for regions very distant from the applicator, where mean dose values are very low, uncertainties are higher and relative differences may reach 20.4%. For the tumor base and the eye structures closest to the applicator, such as sclera, choroid and retina, the maximum difference observed was 4%, presenting the heterogeneous model higher mean dose values. For the other eye regions, the higher doses were obtained when the homogeneous water eye model is taken into consideration. Mean dose distributions determined for the homogeneous water eye model are similar to those obtained for the

  10. Assessment, modelization and analysis of {sup 106} Ru experimental transfers through a freshwater trophic system; Evaluation, modelisation et analyse des transferts experimentaux du {sup 106}Ru au sein d`un reseau trophique d`eau douce

    Vray, F.

    1994-11-24

    Experiments are carried out in order to study {sup 106} RU transfers through a freshwater ecosystem including 2 abiotic compartments (water and sediment) and 3 trophic levels (10 species). Experimental results are expressed mathematically so as they can be included into a global model which is then tested in two different situations. The comparison of the available data concerning the in situ measured concentrations to the corresponding calculated ones validates the whole procedure. Analysis of the so validated results lightens ruthenium distribution process in the environment. The rare detection of this radionuclide in organisms living in areas contaminated by known meaningful releases can be explained by a relativity high detection limit and by a slight role of the sediment as a secondary contamination source. (author). 78 figs., 18 tabs.

  11. Responses of different dosemeters in beta dosimetry of {sup 106}Ru/{sup 106}Rh ophthalmic applicators;Respostas de diferentes dosimetros termoluminescentes na dosimetria beta de aplicadores oftalmicos de {sup 106}Ru/{sup 106}Rh

    Ferreira, D.F.P.; Daros, K.A.C.; Segreto, R.A.; Medeiros, R.B. [Universidade Federal de Sao Paulo (UNIFESP), Sao Paulo, SP (Brazil)

    2009-07-01

    This work presents the TL response of three kinds of dosimeters from different manufacturing characteristics under irradiation of 106 Ru / 106 Rh sealed sources used in ophthalmic brachytherapy. They are: Ca SO{sub 4}:Dy + teflon (D- Ca SO{sub 4}:Dy -0,4), LiF:Mg, Ti (TLD-100) and Ca SO{sub 4}:Dy (TLD-900). Some of reports accepted by scientific community (NCS report 14 e ICRU report 72) as reference in the quality control of beta applicators dosimetry recommend that the absorbed dose standard uncertainties can be kept below 20%. The TLD Ca SO{sub 4}:Dy + teflon presented proper sensibility and high precision comparing with the others. Considering the similar dimensions of ophthalmic tumors and aside critical structures it is relevant to reduce undesirable effects due to the irradiation of these structures. Therefore, the quality control in the beta dosimetry using this kind of source is a constant challenge. (author)

  12. Process for the removal of 106Ru traces from NH4NO3 effluent generated during recycling of sintered depleted uranium fuel pellets

    Several chemical treatment formulations were tested for the effective removal of very low levels of 106Ru activity from NH4NO3 effluent generated during wet processing of rejected sintered depleted uranium (DU) fuel pellets. Based on the results, a simple process involving precipitation of cobalt sulphide along with ferric hydroxide was selected and further optimization of process variables was carried out. The optimized process has been found to be highly efficient in reducing 106Ru activity down to extremely low levels. (author)

  13. Separation and purification of 106Ru from effluent streams of ion exchange cycle used for Pu purification in PUREX process

    Present paper describes the separation and purification of extracted fraction of Ru using separation techniques namely solvent extraction and extraction chromatography. The feed solution used is ion exchange effluent solution collected from Plant that contained 106Ru activity of ∼ 100 mCi/L level along with 95Nb (∼ 0.29 mCi/L) and Pu (∼ 1.5mg/L) at 7.1 M HNO3. In the initial step, the feed solution is contacted ice with 30% TBP in n-dodecane at organic to aqueous phase ratio of 2:1. The raffinate from this step shows that the free acidity of the solution is reduced from 7.1 to 4.2 M without loosing the Ru activity in the feed

  14. Energy deposition by a 106Ru/106Rh eye applicator simulated using LEPTS, a low-energy particle track simulation

    The present study introduces LEPTS, an event-by-event Monte Carlo programme, for simulating an ophthalmic 106Ru/106Rh applicator relevant in brachytherapy of ocular tumours. The distinctive characteristics of this code are the underlying radiation-matter interaction models that distinguish elastic and several kinds of inelastic collisions, as well as the use of mostly experimental input data. Special emphasis is placed on the treatment of low-energy electrons for generally being responsible for the deposition of a large portion of the total energy imparted to matter. - Highlights: → We present the Monte Carlo code LEPTS, a low-energy particle track simulation. → Carefully selected input data from 10 keV to 1 eV. → Application to an electron emitting Ru-106/Rh-106 plaque used in brachytherapy.

  15. Contamination of Chinese cabbage with 85Sr, 103Ru and 134Cs related to time of foliar application

    A solution containing 85Sr, 103Ru and 134Cs was applied to Chinese cabbage in a greenhouse via foliar spraying at 5 different times during its growth. Interception of the applied activity by plant showed no difference among radionuclides and increased with decreasing time interval between application and harvest. The maximum interception factor observed was 0.87. Percentages of the intercepted activity remaining in the whole leaves at harvest varied 16-58 % for 85Sr, 15-73 % for 103Ru and 33-64 % for 134Cs, with application time and those for the inner leaves (without 6 outmost leaves) varied 2-35 %, 0.4-46 % and 14-40 %, respectively. It was demonstrated that rain plays an important role in weathering loss of the activity. Tying the upper end of the plant prior to the last application lowered interception and remaining activity in the inner leaves by factors of 3-4. Present results can be referred to in predicting the radionuclide concentration in Chinese cabbage and deciding counter-measures at the time of an accidental release from the nuclear installation

  16. Multidimensional dosimetry of {sup 106}Ru eye plaques using EBT3 films and its impact on treatment planning

    Heilemann, G., E-mail: gerd.heilemann@meduniwien.ac.at; Kostiukhina, N. [Department of Radiation Oncology/Comprehensive Cancer Center, Medical University of Vienna/AKH Vienna, Vienna 1090 (Austria); Nesvacil, N.; Georg, D. [Department of Radiation Oncology/Comprehensive Cancer Center, Medical University of Vienna/AKH Vienna, Vienna 1090, Austria and Christian Doppler Laboratory for Medical Radiation Research for Radiation Oncology, Vienna 1090 (Austria); Blaickner, M. [Health and Environment Department Biomedical Systems, Austrian Institute of Technology GmbH, Vienna 1220 (Austria)

    2015-10-15

    Purpose: The purpose of this study was to establish a method to perform multidimensional radiochromic film measurements of {sup 106}Ru plaques and to benchmark the resulting dose distributions against Monte Carlo simulations (MC), microdiamond, and diode measurements. Methods: Absolute dose rates and relative dose distributions in multiple planes were determined for three different plaque models (CCB, CCA, and COB), and three different plaques per model, using EBT3 films in an in-house developed polystyrene phantom and the MCNP6 MC code. Dose difference maps were generated to analyze interplaque variations for a specific type, and for comparing measurements against MC simulations. Furthermore, dose distributions were validated against values specified by the manufacturer (BEBIG) and microdiamond and diode measurements in a water scanning phantom. Radial profiles were assessed and used to estimate dosimetric margins for a given combination of representative tumor geometry and plaque size. Results: Absolute dose rates at a reference depth of 2 mm on the central axis of the plaque show an agreement better than 5% (10%) when comparing film measurements (MCNP6) to the manufacturer’s data. The reproducibility of depth-dose profile measurements was <7% (2 SD) for all investigated detectors and plaque types. Dose difference maps revealed minor interplaque deviations for a specific plaque type due to inhomogeneities of the active layer. The evaluation of dosimetric margins showed that for a majority of the investigated cases, the tumor was not completely covered by the 100% isodose prescribed to the tumor apex if the difference between geometrical plaque size and tumor base ≤4 mm. Conclusions: EBT3 film dosimetry in an in-house developed phantom was successfully used to characterize the dosimetric properties of different {sup 106}Ru plaque models. The film measurements were validated against MC calculations and other experimental methods and showed a good agreement with

  17. Biochemistry of derivatives of amino acid with (/sup 103/Ru)ruthenocene. Comparison with /sup 131/I-hippuran

    Wenzel, M.; Park, I.-H.

    1986-01-01

    The potential radiopharmaceuticals: ruthenocenoyl alanine, ruthenocenoyl methionine, 1'methyl-ruthenocenoyl glycine and its esters were labelled with /sup 103/Ru starting from the analogous ferrocene compounds. In a series of tests in mice and rats these substances were compared with hippuran and ruppuran (=ruthernocenoyl glycine, a ruthenocene-amino acid analogue of hippuran). The organ distribution of these compounds was measured at various times after injection. Kidney concentrations of 1'-methyl-ruthenocenoyl glycine and its esters were found to be extremely high, followed by a rapid excretion. In contrast with these compounds, ruthenocenoyl methionine indicated a significantly greater affinity for liver than for kidney, but not for pancreas. Ruthenocenoyl alanine exhibits a high affinity for tumor cells. The advantages of /sup 97/Ru labelled radiopharmaceuticals compared with sup(99m)Tc or /sup 123/I//sup 131/I labelled compounds are discussed.

  18. Determination of 106Ru, 134/137Cs, and 241Am concentrations and Action Level in the Foodstuffs Consumed by Inhabitants of Iraq

    *H. N. Majeed

    2013-03-01

    Full Text Available The specific activity concentrations of (106Ru, 134/137Cs, and 241Am nuclides in 40 imported foodstuffs which collected randomly in January 2012 from all Iraqi cities markets were studied. The rang of specific activity concentrations of 106Ru varies from (37.930±6.16 Bq kg-1 (S No. :17: Turkey Kidney bean to 99.735±9.99 Bq kg-1 (S No.:32: Egypt Broad bean, with average value 71.667±8.47 Bq kg-1. For 134Cs varies from 0.200±0.45 Bq kg-1 (S No. :19 : Ukraine Chick-pea to 2.365±1.54 Bq kg-1 (S No. :33 : Peru Broad bean with average value (0.988±0.99 Bq kg-1.The activity concentrations of 137Cs varies from 0.164±0.40 Bq kg-1 (S No.:19 : Ukraine Chick-pea to 5.291±2.30 Bq kg-1 ( S No.: 39: Uzbekistan Mung bean with average value 1.460±1.21, then for 241Am the activity concentrations varies from 0.029±0.17 Bq kg-1 (S No.:23 : Iran Chick-pea to 1.248±1.12 Bq kg-1 (S No.:40: Canada Green peas with average value 0.399±0.63. All the values were less than the World average concentrations [15,17]. The high contributor for 106Ru, 134/137Cs, and 241Am radionuclides were in Broad bean and other foodstuffs (which contained Brown grit, White grit, Mung bean and Green peas as a 12%, Broad bean as 14%, corn as a 19% and other foodstuffs with 15% respectively The lowest contributor of 106Ru, 134/137Cs, and 241Am radionuclides in the studied foodstuffs were 6% in cowpea, 7% in semolina, 5% in lentil and 4% in lentil respectively. The action level of the 106Ru, 134/137Cs, and 241Am radionuclide’s for three age groups have been calculated and the foodstuffs were within the range permitted and free of any radiation and thus there was no seriousness in dealing with.

  19. Nondestructive analysis of RA reactor fuel burnup, Program for burnup calculation base on relative yield of 106Ru, 134Cs and 137Cs in the irradiated fuel

    Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of 106Ru, 134Cs and 137Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values

  20. Study of Relationship Between Neutron Energy and Fission Yields of 95Zr, 140Ba and 147Nd From 235U

    2001-01-01

    This work measures fission yields of 235U induced by neutrons with energy of thermal, 3.0, 5.0, 5.5, 8.0 and 14.8 MeV. The main purpose is to study the relationship between neutron energy and fission fields of 95Zr,140Ba and 147Nd from 235U by measuring the radioactivity of foil with direct gamma spectrometry. The fission yields induced by fast neutrons are get by fast-thermal-ratio method which based on yields from thermal neutrons, yields by thermal neutron are come from absolute measurement. Since fast-thermal-ratio method eliminates uncertainties of gamma intensity, gamma

  1. Mean retention time of 51Cr-EDTA and 103Ru-phenanthroline in the digestive tract of sheep and bulls after feeding on straw pellets

    Two lots of straw pellets (supplemented with 10% molasses), produced either with a 5 mm sieve in a hammer mill (lot A) or with a 12 mm sieve (lot B) from wheat straw, were tested with 4 sheep (wethers, average live weight 43 kg) and 4 bulls (average live weight 170 kg). After carrying out a digestibility experiment, the mean retention time, the 80% excretion of the markers and the transit time were ascertained with the help of 51Cr-EDTA and 103Ru-phenanthroline. The digestibility of carbohydrates (both crude fiber and N-free extractives) was significantly higher for the bulls than for the sheep. (author)

  2. 2. Quarterly progress report, 1978

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  3. 1. Quaterly progress report, 1979

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  4. 2. Quarterly progress report, 1977

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  5. Monthly progress report

    This monthly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts (air at ground level, high altitude air), rainwater, surface water, underground water, irrigation water, drinking water, food chain (milk, plants, cattle, fish), sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A monthly bibliographic selection is also presented

  6. 4. Quarterly progress report, 1978

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  7. 2. Quarterly progress report, 1981

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radiosotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  8. Monthly progress report

    This monthly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts(air at ground level, high altitude air), rainwater, surface water, underground water, irrigation water, drinking water, food chain (milk, plants, cattle, fish), sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A monthly bibliographic selection is also presented

  9. Monthly progress report

    This monthly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts (air at ground level, high altitude air), rainwater, surface water, underground water, irrigation water, drinking water, food chain (milk, plants, cattle, fish), sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 104Rh, 226Ra, 54Mn, U and T). A monthly bibliographic selection is also presented

  10. 3. Quarterly progress report, 1981

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 95Zr and 95Nb, 103Ru, 131I, 137Cs, 140Ba and 140La, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  11. Energy deposition by a {sup 106}Ru/{sup 106}Rh eye applicator simulated using LEPTS, a low-energy particle track simulation

    Fuss, M.C. [Instituto de Fisica Fundamental, Consejo Superior de Investigaciones Cientificas (CSIC), Serrano 113-bis, 28006 Madrid (Spain); Munoz, A.; Oller, J.C. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Avenida Complutense 22, 28040 Madrid (Spain); Blanco, F. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad Complutense de Madrid, Avenida Complutense, 28040 Madrid (Spain); Williart, A. [Departamento de Fisica de los Materiales, Universidad Nacional de Educacion a Distancia, Senda del Rey 9, 28040 Madrid (Spain); Limao-Vieira, P. [Laboratorio de Colisoes Atomicas e Moleculares, Departamento de Fisica, CEFITEC, FCT-Universidade Nova de Lisboa, Quinta da Torre, 2829-516 Caparica (Portugal); Borge, M.J.G.; Tengblad, O. [Instituto de Estructura de la Materia, Consejo Superior de Investigaciones Cientificas (CSIC), Serrano 113-bis, 28006 Madrid (Spain); Huerga, C.; Tellez, M. [Hospital Universitario La Paz, Paseo de la Castellana 261, 28046 Madrid (Spain); Garcia, G., E-mail: g.garcia@iff.csic.es [Instituto de Fisica Fundamental, Consejo Superior de Investigaciones Cientificas (CSIC), Serrano 113-bis, 28006 Madrid (Spain); Departamento de Fisica de los Materiales, Universidad Nacional de Educacion a Distancia, Senda del Rey 9, 28040 Madrid (Spain)

    2011-09-15

    The present study introduces LEPTS, an event-by-event Monte Carlo programme, for simulating an ophthalmic {sup 106}Ru/{sup 106}Rh applicator relevant in brachytherapy of ocular tumours. The distinctive characteristics of this code are the underlying radiation-matter interaction models that distinguish elastic and several kinds of inelastic collisions, as well as the use of mostly experimental input data. Special emphasis is placed on the treatment of low-energy electrons for generally being responsible for the deposition of a large portion of the total energy imparted to matter. - Highlights: > We present the Monte Carlo code LEPTS, a low-energy particle track simulation. > Carefully selected input data from 10 keV to 1 eV. > Application to an electron emitting Ru-106/Rh-106 plaque used in brachytherapy.

  12. Electrical conductivity spectra of Sn doped BaTi0.95Zr0.05O3

    The alternating current (ac) conductivity spectra of Sn doped BaTi0.95Zr0.05O3 prepared by solid state reaction have been studied in the temperature range of 373–473 K. Mixed valency of Sn atoms and the oxygen vacancy controls electrical transport process. The ac conductivity follows Jonscher type power law as a function of frequency. Derived dc conductivity and hopping frequency follow Arrhenius type temperature dependency and have same activation energy. Almost temperature independent nature of frequency exponent indicates that the electrical conduction in Zr and Sn co-doped BaTiO3 relaxor is quantum mechanical electron tunneling. The conductivity spectra are perfectly scaled using the scaling parameters as dc conductivity and hopping frequency.

  13. Generalized loss of particle and hole neutron spectroscopic intensity in 103 Ru: a sign of significant differences between the odd nucleus and the G.S. of both even neighbors

    New spectroscopic strength information extracted through the 102 Ru (d,p) 103 Ru reaction with the nuclear emulsion technique at the Sao Paulo Pelletron - Enge spectrograph, significantly extended the range and detail of previously available data. The particle strength detected, as has already been verified for the hole strength, is also considerably depleted with respect to the valence expectations, in contrast with what was obtained for 101 Ru. This fact is interpreted as a sign of significant differences between the states of 103 Ru and the G.S. of both even neighbors. (author)

  14. Generalized loss of particle and hole neutron spectroscopic intensity in {sup 103} Ru: a sign of significant differences between the odd nucleus and the G.S. of both even neighbors

    Barbosa, M.D.L.; Duarte, J.M.; Borello-Lewin, T.; Ukita, G.M.; Gomes, L.C.; Horodynski-Matsushigue, L.B. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    1997-12-31

    New spectroscopic strength information extracted through the {sup 102} Ru (d,p) {sup 103} Ru reaction with the nuclear emulsion technique at the Sao Paulo Pelletron - Enge spectrograph, significantly extended the range and detail of previously available data. The particle strength detected, as has already been verified for the hole strength, is also considerably depleted with respect to the valence expectations, in contrast with what was obtained for {sup 101} Ru. This fact is interpreted as a sign of significant differences between the states of {sup 103} Ru and the G.S. of both even neighbors. (author) 3 refs., 1 fig.; mbarbosa at if.usp.br

  15. An experimental study of the time dependence of uptake from soil of 137Cs, 106Ru, 144Ce and 99Tc into green vegetables, wheat and potatoes

    In this study the experimental data were analysed using the CEGB's dynamic foodchain model, and were used to validate the relevant part of the model structure, to produce model-specific input data and to identify possible future improvements to the model structure. The root uptake of the specified radionuclides was studied and the concentration levels measured. The data were analysed using a simplified version of the general model. The compartment system incorporated within the model was shown to be capable of reproducing the data for 137Cs, 106Ru and 144Ce to an extent sufficient to justify its use in ingestion radiological dose assessments, but to be less successful in fitting the 99Tc data. The analysis resulted in the production of a well validated set of model-specific input data relevant to UK conditions and agricultural practice differing significantly from values obtained from global literature surveys. Possible future improvements to the model structure were also identified, aimed at providing improved estimates of crop contamination levels for timescales in excess of those considered in this study. (U.K.)

  16. Does escalation of the apical dose change treatment outcome in β-radiation of posterior choroidal melanomas with 106Ru plaques?

    Purpose: To show the results of treating posterior uveal melanomas with 106Ru plaque β-ray radiotherapy and to review and discuss the literature concerning the optimal apical dose prescription (100 vs. 160 Gy). Methods and Materials: Forty-eight patients with uveal melanomas (median height 3.85 mm + 1 mm sclera) were treated with ruthenium plaques. The median apical dose was 120 Gy, the median scleral dose 546 Gy. Results: After 5.8 years of follow-up, the overall 5-year survival rate was 90%, the disease specific 5-year survival rate was 92% (3 patients alive with metastasis). Six percent received a second ruthenium application, 10% of the eyes had to be enucleated. Local control was achieved in 90% of the patients with conservative therapy alone. Central or paracentral tumors showed 50% of the pretherapeutic vision after 4 years, and 80% of the vision was preserved in those with peripheral tumors. The main side effects were mostly an uncomplicated retinopathy (30%); macular degeneration or scarring led to poor central vision in 30% of cases. Conclusion: Brachytherapy with ruthenium applicators is an effective therapy for small- and medium-size posterior uveal melanomas. Our results are comparable to other series. The treatment outcome does not seem to be capable of improvement by increasing the apical dose. An internationally accepted model for defining the dosage in brachytherapy is needed

  17. Studies on treatment of low level radioactive liquid waste for removal of anionic species of 125Sb, 99Tc and 106Ru. Contributed Paper RD-14

    The treatment of intermediate level waste at Waste Immobilization Plant generates low level radioactive waste which would require further management before discharge to sea. This waste is expected to contain polymeric oxo anions of 125Sb, 99Tc, 106Ru in addition to cationic species like 137Cs, 90Sr etc. Chemical treatment takes care of the major contributors to radioactivity viz 137Cs, 90Sr etc but traces of activity due to anionic species remain in the treated waste effluent. Novel composite anionic exchanger namely Polyurethane foam coated with Hydrous Zirconium Oxide was developed for removal of these anionic species. This material was successfully employed for removal of anionic 125Sb from radioactive waste effluent at Waste Management Division, Trombay. Based on our experience with Sb removal using the above material it was decided to assess the ability of the exchanger in removal of other anionic species bearing Ru and Tc. It was observed that in addition to complete removal of Sb, 50% Ru removal and 40% Tc could also be removed using this material from radioactive waste effluents. In lab experiments, similar results were obtained with simulated low level waste bearing inactive Ru. Among several hydrous oxides tried in a batch study, Hydrous Zirconium Oxide showed a maximum removal of 40% for Tc in actual waste generated from reprocessing plant. Based on the above it has been planned to set up an anion exchange column with Hydrous Zirconium Oxide coated Polyurethane foam for final treatment of chemically treated waste effluent prior to discharge as a prime step towards achieving our goal of minimum discharge to Sea. (author)

  18. XAFS study of an intermetallic TiFe0.95Zr0.03Mo0.02 system for CO2 conversion

    An intermetallic TiFe0.95Zr0.03Mo0.02 system used for CO2 conversion was studied by XAFS spectroscopy. The initial samples prepared by the skull-furnace method were exposed to the below treatments: (1) H2 absorption (up to ∼1 mol of H2 per 1 mol of the initial sample); (2) H2 thermal desorption (up to ∼0.18 mol of H2 at 180 deg. C in inert argon atmosphere); (3) hydrogen titration by CO2 (at 350 deg. C for 4-6 h). For samples [TiFe0.95Zr0.03Mo0.02]H2x (x≤0.18), the position of Ti-K edge shifts of about ∼3 eV compared to that of the initial sample. From the XAFS studies of the modified [TiFe0.95Zr0.03Mo0.02] intermetallide, it is established that the interaction of dissolved hydrogen and titanium follows the donor-acceptor mechanism. Hydrogen atoms are located on the straight line connecting two titanium atoms. The one distance between a hydrogen atom and a titanium atom is less than 1.5 A. It is proposed that the effect of strong binding of hydrogen in the [TiFe0.95Zr0.03Mo0.02]H2x intermetallide at a molar concentration of absorbed hydrogen of 0.18 is associated with the formation of such binding

  19. Studies of 131I, 137Cs and 103Ru in milk, meat and vegetables in North East Scotland following the Chernobyl accident

    Uptake and clearance of radionuclides in foodstuffs have been studied in the neighbourhood of Aberdeen in North East Scotland following the Chernobyl accident. The level of 131I in goats' milk was 100-200 Bq litre-1 in early May and declined with an effective half-life of 4.3 days, but that in cows's milk was only a few Bq litre-1 as most cattle were kept indoors. 137Cs and 103Ru activities in broccoli declined with effective half-lives of 11 and 6 days respectively, while 137Cs in grass decreased with a half-life of 22 days, the reduction appearing to show a relationship to weekly rainfall. Studies of tissues from groups of lambs initially grazed on contaminated pasture and later (a) fed indoors on concentrates or (b) continuing to graze outdoors, showed the 137Cs concentrations to decline with half-lives of (a) 17 days and (b) 25 days, while the half-lives describing the reduction in total 137Cs activity were (a) 20 days and (b) 35 days. (author)

  20. Determination of absorbed dose distribution in water for COC ophthalmic applicator of {sup 106}Ru/{sup 106}Rh using Monte Carlo code-MCNPX; Determinacao da distribuicao de dose absorvida na agua para o aplicador oftalmico COC de {sup 106}Ru/{sup 106}Rh utilizando o codigo de Monte Carlo - MCNPX

    Barbosa, Nilseia A.; Rosa, Luiz A. Ribeiro da, E-mail: nilseia@ird.gov.br, E-mail: lrosa@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ),Rio de Janeiro, RJ (Brazil); Braz, Delson, E-mail: delson@nuclear.ufrj.br [Coordenacao dos programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2014-07-01

    The COC ophthalmic applicators using beta radiation source of {sup 106}Ru/{sup 106}Rh are used in the treatment of intraocular tumors near the optic nerve. In this type of treatment is very important to know the dose distribution in order to provide the best possible delivery of prescribed dose to the tumor, preserves the optic nerve region extremely critical, that if damaged, can compromise the patient's visual acuity, and cause brain sequelae. These dose distributions are complex and doctors, who will have the responsibility on the therapy, only have the source calibration certificate provided by the manufacturer Eckert and Ziegler BEBIG GmbH. These certificates provide 10 absorbed dose values at water depth along the central axis applicator with the uncertainties of the order of 20% isodose and in a plane located 1 mm from the applicator surface. Thus, it is important to know with more detail and precision the dose distributions in water generated by such applicators. To this end, the Monte Carlo simulation was used using MCNPX code. Initially, was validated the simulation by comparing the obtained results to the central axis of the applicator with those provided by the certificate. The different percentages were lower than 5%, validating the used method. Lateral dose profile was calculated for 6 different depths in intervals of 1 mm and the dose rates in mGy.min{sup -1} for the same depths.

  1. Preparation and characterization of Ce0.95Zr0.05O2 nanopowders obtained by sol–gel and template methods

    Ce0.95Zr0.05O2 nanopowders have been prepared by a standard Pechini-type sol–gel process and by using polymethyl methacrylate (PMMA) colloidal crystals as template. The effects of these different synthesis routes, on the structure and microstructural features of the nanopowders, were evaluated by using X-ray diffraction (XRD), scanning electron microscopy (SEM) and specific surface area measurements. For both preparation routes, the XRD analysis has shown that a cubic fluorite structure is formed with a crystallite size of ∼45–50 nm. The SEM images indicate that the powder obtained by the sol–gel Pechini-type process, is constituted by aggregated nanoparticles with relatively uniform shape and size, whereas the powder synthesized as inverse opal exhibits the formation of macropores with a mean size of ∼130 nm. The specific surface areas of the powder samples obtained by the Pechini-type sol–gel and inverse opal methods are ∼56 m2 g−1 and ∼90 m2 g−1 respectively. Additionally, the thermoluminescence (TL) signal of the synthetized samples has been measured in order to examine its potential application in the field of dosimetry of ionizing radiations. -- Highlights: ► Ce0.95Zr0.05O2 nanopowders synthesized by sol–gel and template methods. ► Influence of the synthesis method on structural and microstructural features. ► Analysis of the thermoluminescence response. ► Study of application on dosimetry field.

  2. Radioactivity in surface air and precipitation in Japan after the Chernobyl accident

    Radioactive plumes from the Chernobyl reactor accident first passed over Japan on 3 May 1986. Measurements of 103Ru, 131I and 137Cs in rainfall and airborne dust collected at Chiba near Tokyo show that, in fact, at least two or more kinds of plume arrived during May. Their altitudes were calculated to be about 1500 m in early May and 6300 m in late May. Radionuclides detected in 33 precipitation samples collected by the network of radiation monitoring stations from 1 to 22 May were 7Be, 89Sr, 90Sr, 95Zr, 95Nb, 103Ru, 106Ru, sup(110m)Ag, 125Sb, sup(129m)Te, 131I, 132Te, 132I, 134Cs, 136Cs, 137Cs, 140Ba, 140La, 141Ce and 144Ce. The radiation was characterized by higher levels of the volatile nuclides, such as 103Ru, 132Te, 131I and 137Cs, than fallout levels in nuclear weapons testing, and by activity ratios of 0.48 and 14 for, respectively, 134Cs/137Cs and 89Sr/90Sr, as on 26 April. the fallout activity was higher in Northwestern Japan, the average depositions of 90Sr and 137Cs in Japan from 1 May (or 30 April) to 22 May being 1.4 Bq m-2 and 95 Bq m-2, inventories which are 14 and 550 times higher than the pre-Chernobyl values. (author)

  3. Evolution of gamma artificial radioactivity in coastal sediments of the English Channel during the years 1976, 1977 and 1978

    During 1976-1977, a state of equilibrium was found to prevail for 106Ru and 144Ce, especially in the North-West Cotentin and the Norman-Breton gulf, where reconcentration of both radionuclides was observed with preferential enrichment of the latter over the former. Levels of 125Sb and 137Cs were found to be low but were difficult to interpret, because of the particular physico-chemical behavior of 125Sb and the long half-life of 137Cs. The results obtained for 103Ru, 141Ce, 95Zr may be explained entirely by the contribution of atmospheric fallout. 144Ce and 106Ru levels in the Norman-Breton gulf may be for the most part ascribed to La Hague disposals, radionuclide dispersal from the emissary being characterized by an eastward transfer of the soluble fraction and a westward transfer of the particulate fraction, with transit times which may last up to 2 years. The boundary between the areas submitted respectively to the twofold impact of fallout and industrial waste, and to fallout alone would appear to lie between the mouth of the Trieux river and Morlaix Bay. From a graphic representation of the relationship between radionuclides, empiric distribution laws for 106Ru and 137Cs were established from 144Ce level parameters characteristic of the areas considered (years 1976-1977)

  4. LLX separation of carrier-free, 94,95,97,103Ru, 93,94,95,96,99mTc and 95,96Nb produced in alpha-particle activated molybdenum by TOA

    A radiochemical charged particle activation procedure for the simultaneous production of carrier-free radioisotopes of more than one element in a single target and their subsequent separation through LLX has been demonstrated. The carrier free isotopes, 95,96Nb, 93,94,95,96,99mTc and 94,95,97,103Ru formed through Mo(α,αpxn), Mo(α,pxn) and Mo(α,xn) nuclear reactions with 40 MeV α-particle as detected by nondestructive γ-ray spectroscopy, have been effectively separated through LLX using TOA as an anionic extractant. Separation of the bulk matrix of molybdenum from the carrier free products has been monitored radiometrically using isotopic 93,99mMo formed through the Mo(α,αxn) reaction, as radioindicators for the target element. Purity of the separated carrier free radionuclide has been verified by γ-ray spectrometry. (author)

  5. Quantitative analysis of fission products by γ spectrography

    The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio (144Ce + 144Pr activity)/ 137Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By γ-scintillation spectrography it was possible to estimate the following elements individually: 141Ce, 144Ce + 144Pr, 103Ru, 106Ru + 106Rh, 137Cs, 95Zr + 95Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author)

  6. Calculated activities of some isotopes in the RA reactor highly enriched fuel significant for possible environmental contamination - Operational report

    This report contains calculation basis and obtained results of activities for three groups of isotopes in the RA reactor 80% enriched fuel element. The following isotopes are included: 1) 85mKr, 87Kr, 88Kr, 131J, 132J, 133J, 134J, 135J, 133Xe, 138Xe i 138Cs, 2) 89Sr, 90Sr, 91Sr, 92Sr, 95Zr, 97Zr, 103Ru, 105Ru, 106Ru, 129mTe, 134Cs, 137Cs, 140Ba, 144Ce, kao i 3) 238Pu, 239Pu i 240Pu. It was estimated that the fuel is exposed to mean neutron flux. The periodicity of reactor operation is taken into account. Calculation results are given dependent on the time of exposure. These results are to be used as source data for Ra reactor safety analyses

  7. Application of the radiochemical - and the direct gamma ray spectrometry method to the burnup determination of irradiated uranium oxide

    The burn up of natural U3O8 that occurs by the action of thermal neutrons was determined, using the radioisotopes 144Ce, 137Cs, 103Ru, 106Ru and 95Zr as monitors. The determination of the burn up was made using both destructive and non-destructive methods. In the non-destructive method, the technique of direct gamma-ray spectrometry was used and the radioisotopes mentioned were simultaneously counted in a Ge-Li detector. In the radiochemical method the same radioisotopes were isolated one from the other and from all other fission products before counting. The solvent extraction technique was used for the radiochemical separation of uranium, cerium, cesium and ruthenium. To separate zirconium and niobium, adsorption in silica-gel was used. The extraction agent employed to isolate cesium was dipycrilamine and for the separation of the other radioisotopes Di-(2-Ethyl Hexyl) Phosphoric acid (HDEHP) was used. (Author)

  8. Assessment of selected fission products in the Savannah River Site environment

    Most of the radioactivity produced by the operation of a nuclear reactor results from the fission process, during which the nucleus of a fissionable atom (such as 235U) splits into two or more nuclei, which typically are radioactive. The Radionuclide Assessment Program (RAP) has reported on fission products cesium, strontium, iodine, and technetium. Many other radionuclides are produced by the fission process. Releases of several additional fission products that result in dose to the offsite population are discussed in this publication. They are 95Zr, 95Nb, 103Ru, 106Ru, 141Ce, and 144Ce. This document will discuss the production, release, migration, and dose to humans for each of these selected fission products

  9. Microwave property improvement of Ca[(Li1/3Nb2/3)0.95Zr0.15]O3+δ perovskite by A-site substitution

    Hu, Mingzhe; Xiong, Gang; Ding, Zhao

    2016-04-01

    The crystal structure and microwave dielectric properties of Ca[(Li1/3Nb2/3)0.95Zr0.15]O3+δ ceramic (CLNZ) are tuned by A-site substitution of Sr2+ and Ba2+ ions in the present paper. The tuning effect on the crystal structure is investigated by the X-ray diffraction (XRD) pattern and it illustrates that single phase of orthorhombic perovskite structure is formed, however, minor amount of BaNb2O6-type second phase is also detected in (Ca1‑xBax)[(Li1/3Nb2/3)0.95Zr0.15]O3+δ ceramics (CBLNZ) in the range of x ≥ 0.025, while pure perovskite phase is obtained in (Ca1‑xSrx)[(Li1/3Nb2/3)0.95Zr0.15]O3+δ ceramics (CSLNZ) in the whole investigation range of 0 ≤ x ≤ 0.2. With the increase of x value, the unit cell volumes of both CBLNZ and CSLNZ perovskites gradually expand, which results in the degradation of the vibration bond strength between the B-site ions and oxygen in the perovskites. The microscopic structure related thermal parameters in CSLNZ and CBLNZ perovskites are analyzed in terms of Clausius-Mossotti equation to reveal the original contributors in the temperature coefficients. The results show that both Sr2+ and Ba2+ substitution can effectively improve the permittivity and Qf value, especially, improve the temperature coefficient of CLNZ ceramic in a certain range.

  10. Application of hydrous ceria for the uptake of several cations and in the separation of carrier-free 99mTc and 95Nb from 99Mo and 95Zr respectively

    Hydrous ceria of suitable column quality has been prepared by precipitation with ammonia from a solution of ceric sulphate in hot dil. H2SO4, and subsequent drying at 70degC. The composition of the material has been ascertained by TGA to be CeO2.2.2H2O. The IEC (ion exchange capacity) value is found to be 0.25 meq/g. Uptake of 20 tracer cations has been studied on this exchanger. The data show that sup(55+59)Fe, 99Y, rare earth isotopes, 95Zr, 125Sb, 99Mo and 185W are adsorbed in the exchanger in an appreciable manner. Radiochemical separations of 99mTc and 95Nb respectively from 99Mo and 95Zr have been carried out by using a very simple chemical procedure over the columns of freshly prepared ceria material. γ-Ray spectra of the separated 99mTc and 95Nb show that the products are of high radionuclidic purity. Overall separation procedures are quick and give quantitative yield. (author). 21 refs., 3 figs., 1 tab

  11. Preparation and characterization of Ce{sub 0.95}Zr{sub 0.05}O{sub 2} nanopowders obtained by sol-gel and template methods

    Isasi, J., E-mail: isasi@quim.ucm.es [Departamento de Quimica Inorganica I, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Ciudad Universitaria s/n, 28040 Madrid (Spain); Perez, M. [Departamento de Quimica Inorganica I, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Ciudad Universitaria s/n, 28040 Madrid (Spain); Castillo, J.F. [Departamento de Quimica Fisica I, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Ciudad Universitaria s/n, 28040 Madrid (Spain); Correcher, V. [CIEMAT, Av. Complutense 22, 28040 Madrid (Spain); Aldama, I.; Arevalo, P.; Carbajo, M.C. [Departamento de Quimica Inorganica I, Facultad de Ciencias Quimicas, Universidad Complutense de Madrid, Ciudad Universitaria s/n, 28040 Madrid (Spain)

    2012-09-14

    Ce{sub 0.95}Zr{sub 0.05}O{sub 2} nanopowders have been prepared by a standard Pechini-type sol-gel process and by using polymethyl methacrylate (PMMA) colloidal crystals as template. The effects of these different synthesis routes, on the structure and microstructural features of the nanopowders, were evaluated by using X-ray diffraction (XRD), scanning electron microscopy (SEM) and specific surface area measurements. For both preparation routes, the XRD analysis has shown that a cubic fluorite structure is formed with a crystallite size of {approx}45-50 nm. The SEM images indicate that the powder obtained by the sol-gel Pechini-type process, is constituted by aggregated nanoparticles with relatively uniform shape and size, whereas the powder synthesized as inverse opal exhibits the formation of macropores with a mean size of {approx}130 nm. The specific surface areas of the powder samples obtained by the Pechini-type sol-gel and inverse opal methods are {approx}56 m{sup 2} g{sup -1} and {approx}90 m{sup 2} g{sup -1} respectively. Additionally, the thermoluminescence (TL) signal of the synthetized samples has been measured in order to examine its potential application in the field of dosimetry of ionizing radiations. -- Highlights: Black-Right-Pointing-Pointer Ce{sub 0.95}Zr{sub 0.05}O{sub 2} nanopowders synthesized by sol-gel and template methods. Black-Right-Pointing-Pointer Influence of the synthesis method on structural and microstructural features. Black-Right-Pointing-Pointer Analysis of the thermoluminescence response. Black-Right-Pointing-Pointer Study of application on dosimetry field.

  12. Appendix to Health and Safety Laboratory environmental quarterly, March 1, 1976--June 1, 1976. [Tabulated data on content of lead in surface air and /sup 7/Be, /sup 95/Zr, /sup 137/Cs, /sup 144/Ce, and /sup 90/Sr in surface air, milk, drinking water, and foods sampled in USA

    Hardy, E.P. Jr.

    1976-07-01

    Tabulated data are presented on: the monthly deposition of /sup 89/Sr and /sup 90/Sr at some 100 world land sites; the content of lead and /sup 7/Be, /sup 95/Zr, /sup 137/Cs, and /sup 144/Ce in samples of surface air from various world sites; and the content of /sup 90/Sr in samples of milk, drinking water, and animal and human diets collected at various locations throughout the USA. (CH)

  13. Application of radiochemical-and direct gamma ray spectrometry methods for the determination of the burnup of irradiated uranium oxide

    The burn-up of U3O8 (natural uranium) samples was determined by using both destructive and non-destructive methods, and comparing the results obtained. The radioisotopes 144Ce, 103Ru, 106Ru, 137Cs and 95Zr were chosen as monitors. In order to isolate the radioisotopes chosen as monitors, a separation scheme has been established in which the solvent extraction technic is used to separate cerium, cesium, and ruthenium one from the other and from uranium. The separation between zirconium and niobium and of both from the others was accomplished by means of adsorption on a silica-gel column. When the non-destructive method was used, the radioactivity of each nuclide of interest was measured in the presence of all others. For this purpose, use was made of gamma-ray spectrometry and a Ge-Li detector. The comparison of burn-up values obtained by both destructive and non-destructive methods was made by means of Student's 't' test, and it has shown that the averages of results obtained in each case are equal. (Author)

  14. Reconstruction of the composition of the Chernobyl radionuclide fallout and external radiation absorbed doses to the population in areas of Russia

    The results of reconstruction of the radionuclide composition of the Chernobyl fallout in the territories of Russia is presented. Reconstruction has been carried out by means of statistical analysis of the gamma spectrometry data on 2867 soil samples collected in the territories of Ukraine, Byelarus and Russia from 1986 to 1988. To verify the data, aggregated estimates of the fuel composition of the 4th block at the moment of the accident (available from the literature) have been used, as well as the estimates of activity released to the atmosphere. As a result, correlation and regression dependences have been obtained between the activities of the radionuclides most contributing to the dose (137Cs, 134Cs, 131I, 140Ba, 140La, 95Zr, 95Nb, 103Ru, 106Ru, 141Ce, 144Ce, 125Sb). Statistically significant regression relations between different pairs of radionuclides (including analysis of the 'noise' contribution to the data) depending on the distance between the point of sample collection and the power station are presented for the 'north-east track' - the northern part of the 30 km zone and southern part of the Gomel 'district (Byelarus) and the Briansk, Kaluga, Tula and Orel districts (Russia). A methodology is also described for reconstructing space-time characteristics of the contamination of the territories by major dose-forming radionuclides released from the Chernobyl NPP 4th unit. (Author)

  15. Biological effects of radiation: The induction of malignant transformation and programmed cell death

    In the Chernobyl explosions and fire, powderized nuclear fuel was released from the reactor core, causing an unexpected fallout. X-ray analysis and scanning electron microscopy showed that the isolated single particles were essentially pure uranium. These uranium aerosols contained all of the nonvolatile fission products, including the b-emitters, 95Zr, 103Ru, 106Ru, 141Ce, and 144Ce. The hot particles are extremely effective in inducing malignant transformation in mouse fibroblast cells in vitro. The major factor responsible for this effect is focus promotion caused by a wound-mediated permanent increase in cell proliferation (mitogenesis associated with mutagenesis). Transformed foci were analysed for the activation of c-abl, c-erb-A, c-erb-B, c-fms, c-fos, c-myb, c-myc, c-Ha-ras, c-Ki-ras, c-sis, and c-raf oncogenes at the transcriptional level. The pattern of oncogene activation was found to vary from focus to focus. Long interspersed repeated DNA (L1 or LINE makes up a class of mobile genetic elements which can amplify in the cell genome by retroposition. This element is spontaneously transcriptionally activated at a critical population density and later amplified in rat chloroleukaemia cells. UV light and ionizing radiation induce this activation prematurely, and the activation is followed by programmed cell death (apoptosis) in a sequence of events identical to that seen in LIRn activation occurring spontaneously

  16. Burn-up determination of irradiated uranium oxide by means of direct gama spectrometry and by radiochemical method

    The burn-up of thermal neutrons irradiated U3O8 (natural uranium) samples has been determined by using both direct gamma spectrometry and radiochemical methods and the results obtained were compared. The fission products 144Ce, 103Ru, 106Ru, 137Cs and 95Zr were chosen as burn-up monitors. In order to isolate the radioisotopes chosen as monitors, a radiochemical separation procedure has been established, in which the solvent extraction technique was used to separate cerium, cesium and ruthenium one from the other and all of them from uranium. The separation between zirconium and niobium and of both elements from the other radioisotopes and uranium was accomplished by means of adsorption on a silica-gel column, followed by selective elution of zirconium and of niobium. When use was made of the direct gamma-ray spectrometry method, the radioactivity of each nuclide of interest was measured in presence of all others. For this purpose use was made of gamma-ray spectrometry and of a Ge-Li detector. Comparison of burn-up values obtained by both methods was made by means of Student's 't' test, and this showed that results obtained in each case are statistically equal. (Author)

  17. Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

  18. Particle size distribution of radioactive aerosols after the Fukushima and the Chernobyl accidents

    Following the Fukushima accident, a series of aerosol samples were taken between 24th March and 13th April 2011 by cascade impactors in the Czech Republic to obtain the size distribution of 131I, 134Cs, 137Cs, and 7Be aerosols. All distributions could be considered monomodal. The arithmetic means of the activity median aerodynamic diameters (AMADs) for artificial radionuclides and for 7Be were 0.43 and 0.41 μm with GDSs 3.6 and 3.0, respectively. The time course of the AMADs of 134Cs, 137Cs and 7Be in the sampled period showed a slight decrease at a significance level of 0.05, whereas the AMAD pertaining to 131I increased at a significance level of 0.1. Results obtained after the Fukushima accident were compared with results obtained after the Chernobyl accident. The radionuclides released during the Chernobyl accident for which we determined the AMAD fell into two categories: refractory radionuclides (140Ba, 140La 141Ce, 144Ce, 95Zr and 95Nb) and volatile radionuclides (134Cs, 137Cs, 103Ru, 106Ru, 131I, and 132Te). The AMAD of the refractory radionuclides was approximately 3 times higher than the AMAD of the volatile radionuclides; nevertheless, the size distributions for volatile radionuclides having a mean AMAD value of 0.51 μm were very close to the distributions after the Fukushima accident. -- Highlights: • AMADs after the Fukushima and Chernobyl accidents in the Czech Rep. were determined. • The mean value of AMADs of the monitored nuclides from the NPP Fukushima was 0.43 μm. • Nuclides from the NPP Chernobyl fell into two categories – refractory and volatile. • The mean value of AMADs of volatile nuclides from the NPP Chernobyl was 0.51 μm. • AMADs of volatile nucl. from the NPP Chernobyl were 3× smaller than of the refractory radionuclides

  19. Environmental and health consequences in Japan due to the accident at Chernobyl nuclear reactor plant

    A comprehensive review was made on the results of national monitoring program for environmental radioactivity in Japan resulting from the accident at the Chernobyl nuclear power plant in USSR. Period of monitoring efforts covered by the present review is from 30th of April 1986 to 31st of May 1987. A radioactive cloud released from the Chernobyl nuclear reactor initially arrived in Japan on 30th of April 1986 as indicated by the elevated level of 131I, 137Cs and 134Cs activity in the total deposition on 30th of April and also by the increased 137Cs body burden noted on 1st of May. Almost all the radioactive nuclides detected in the European countries were also identified in Japan. For example, the observed nuclides were: 95Zr, 95Nb, 99mTc, 103Ru, 106Ru, 110mAg, 111Ag, 125Sb, 127Sb, 129mTe, 131I, 132Te, 132I, 133I, 134Cs, 136Cs, 137Cs, 140Ba, 140La, 141Ce and 144Ce. Among the above radionuclides, the country average concentration was determined for 131I, 137Cs and 134Cs in various environmental materials such as air, fresh water, soil, milk, leafy and root vegetables, cereals, marine products and other foodstuffs. In contrast to the sharp decline of 131I which was negligible after a few months, 137Cs showed a tendency to maintain its activity in foodstuffs at an appreciable level one year later. Collective effective dose equivalent and dose equivalent to thyroid in Japanese population due to 137Cs, 134Cs and 131I were estimated to be around 590 man Sv and 4760 man Sv, respectively. Corresponding values for the per caput dose equivalent are 5 μSv for whole body and 40 μSv for thyroid, respectively. (author)

  20. Radionuclide accumulation by aquatic biota exposed to contaminated water in artificial ecosystems before and after its passage through the ground

    This study was designed to investigate the comparative accumulation of radionuclides from contaminated water in artificial ecosystems before and after the water's passage through the ground. Fish, clams, algae, and an emergent vascular plant were experimentally exposed to mixtures of radionuclides in three aqueous streams. Two streams consisted of industrial water discharged directly into a leaching trench, and the same water after it had migrated through the ground for a distance of 260 meters. The third stream was river water, which served as a background or control. Biota exposed to river water in the control stream had very low concentrations of 60Co, less than 3 pCi per gram dry weight (pCi/g DW). Other radionuclides were essentially unmeasurable. Biota exposed to trench water accumulated very high relative concentrations of 60Co. Biota exposed to trench water also had measurable concentrations of 155Eu, 144Ce, 141Ce, 125Sb, 124Sb, 103Ru, 106Ru, 137Cs, 95Zr, 95Nb, 58Co, 54Mn, 59Fe, 65Zn, 90Sr, /sup 239,240/Pu, and 238Pu. Biota exposed to ground water had concentrations of 60Co that ranged between 50 and 1200 pCi/g DW. Fish flesh had the lowest concentration of 60Co and algae the highest. Strontium-90 was measured in the tissues of aquatic biota at concentrations ranging between 360 pCi/g DW in clam flesh to 3400 pCi/g DW in leaves and stems of Veronica. Leaves and fruits of tomato plants rooted in the ground water accumulated 90Sr at concentrations of 160 pCi in fruits and 4200 pCi in leaves. Data indicate that 60Co and 90Sr migrated through the ground along with ground-water flow and were available to all classes of aquatic biota and tomato plants rooted in the water via root uptake, sorption, and food chain transfers. 8 refs., 4 figs., 6 tabs

  1. Pollution of atmospheric air with toxic and radioactive particulate matter investigated by means of nuclear techniques

    The application of spectrometric methods of nuclear techniques to the investigations of atmospheric air pollution by toxic and radioactive elements and results of these investigations conducted in the highly industrialized and urbanized regions of Poland have been presented. The method of precipitation of the samples, the measurements and analysis of radiation spectra of alpha and gamma radiation emitted by isotopes present in the samples have been described. The concentrations of toxic metal dust in the air have been evaluated by neutron activation and X-ray fluorescence analysis. Appropriate methods of measurement, calibration of instrument and the discussion of results have been presented. The work presents the results of investigations performed in Siersza within the years 1973-1974 and in Warsaw in the period of 1975-1977, which have permitted to estimate the mean monthly values of concentration in the atmospheric air of the following radioisotopes: 7Be, 54Mn, 95Zr, 103Ru, 106Ru, 125Sb, 131I, 137Cs, 140Ba, 141Ce, 144Ce, 226Ra, Th-nat, U-nat and the following stable elements: Sc, Cr, Fe, Co, Zn, As, Se, Sb, W, Pb. The analysis of changes in concentration of each particular artificial radioisotope in the air for the region of Poland in connection with Chinese nuclear explosions have been given. On the basis of the performed environmental investigations the method of analysis of relations between the concentrations of particular elements present in the dust has been discussed. The applications of this method have been presented. The hazard to the population and the environment caused by the radioactive and toxic dust present in the atmospheric air has been estimated. (author)

  2. Influences of marine sediment on the accumulation of radionuclides by green alga (Ulva pertusa)

    Distribution of radionuclides (60Co, 137Cs, 95Zr-95Nb and 106Ru-106Rh) among green alga (Ulva pertusa), sea water and marine sediment were examined by radioisotope tracer experiment in order to estimate the influence of sediment on the accumulation of radionuclides by the alga. By the application of the compartment model to the experimental results, exponential formulas of distributions were obtained. Through comparison of the transfer coefficients of radionuclides calculated from the exponential formulas, the influence of the sediment on the accumulation of the radionuclides by the green alga was determined to be the largest for 60Co, followed by 95Zr-95Nb, 106Ru-106Rh and 137Cs in this order. The activity ratios of 95Zr-95Nb and 106Ru-106Rh calculated from the transfer coefficients are larger for the alga than for the sediment, inversely those of 60Co and 137Cs show higher values for the sediment than for the alga. Especially, in the case of 60Co, the activity ratio for the sediment is approximately 20 times greater than that for the alga. Biological half lives in green alga estimated from the transfer coefficients were 10 days for 60Co, 7 days for 137Cs, 26 days for 95Zr-95Nb and 24 days for 106Ru-106Rh. (auth.)

  3. Comparison of influences of sediments and sea water on accumulation of radionuclides by worms

    The accumulation and excretion of radionuclides by marine polychaete worms (Nereis japonica) were examined to learn the influence of contaminated sediments on the contamination of marine organisms. The concentration factors of 60Co, 95Zr-95Nb, 106Ru-106Rh and 137Cs for unfed worms were 6, 4, 6 and 6 respectively, and they were similar to the concentration factors for unfed worms. The biological half lives of 60Co, 95Zr-95Nb and 106Ru-106Rh for fed worms were similar to each other (37, 32 and 35 days, respectively) except that of 137Cs (6 days), and all of them were a little shorter than those for unfed worms. The transfer ratios of radionuclides from sediments to worms were 5 per cent for 60Co, 0.9 for 95Zr-95Nb, 0.6 for 106Ru-106Rh and 17.9 for 137Cs in cpm/g in regard to initial activity in sediments. These figures were compared with the concentration factors to estimate the influence of sediments on the contamination of marine organisms. The obtained figures, which we call the biological factor of the sediments, were 120, 440, 1000 and 30 for 60Co, 95Zr-95Nb, 106Ru-106Rh and 137Cs, respectively. (auth.)

  4. Radionuclides contamination of fungi after accident on the Chernobyl NPP

    Zarubina, Nataliia E.; Zarubin, Oleg L. [Institute for Nuclear Research of National Academy of Sciense, 03680, pr-t Nauki, 47, Kiev (Ukraine)

    2014-07-01

    Accumulation of radionuclides by the higher fungi (macromycetes) after the accident on the Chernobyl atomic power plant in 1986 has been studied. Researches were spent in territory of the Chernobyl alienation zone and the Kiev region. Our research has shown that macromycetes accumulate almost all types of radionuclides originating from the accident ({sup 131}I, {sup 140}Ba /{sup 140}La, {sup 103}Ru, {sup 106}Ru, {sup 141}Ce, {sup 144}Ce, {sup 95}Nb, {sup 95}Zr, {sup 137}Cs and {sup 134}Cs). They accumulate the long-living {sup 90}Sr in much smaller (to 3 - 4 orders) quantities than {sup 137}Cs. We have established existence of two stages in accumulation of {sup 137}Cs by higher fungi after the accident on the Chernobyl NPP: the first stage resides in the growth of the concentration, the second - in gradual decrease of levels of specific activity of this radionuclide. Despite reduction of {sup 137}Cs specific activity level, the content of this radionuclide at testing areas of the 5-km zone around the Chernobyl NPP reaches 1,100,000 Bq/kg of fresh weight in 2013. We investigated dynamics of accumulation of Cs-137 in higher fungi of different ecological groups. One of the major factors that influence levels of accumulation of {sup 137}Cs by fungi is their nutritional type (ecological group). Fungi that belong to ecological groups of saprotrophes and xylotrophes accumulate this radionuclide in much smaller quantities than symbio-trophic fungi. As a result of the conducted research it has been established that symbio-trophic fungi store more {sup 137}Cs than any other biological objects in forest ecosystems. Among the symbio-trophic fungi species, species showing the highest level of {sup 137}Cs contamination vary in different periods of time after the deposition. It is connected with variability of quantities of these radio nuclides accessible for absorption at the depth of localization of the main part of mycelium of each species in a soil profile. Soil contamination

  5. [Comparative analysis of the radionuclide composition in fallout after the Chernobyl and the Fukushima accidents].

    Kotenko, K V; Shinkarev, S M; Abramov, Iu V; Granovskaia, E O; Iatsenko, V N; Gavrilin, Iu I; Margulis, U Ia; Garetskaia, O S; Imanaka, T; Khoshi, M

    2012-01-01

    The nuclear accident occurred at Fukushima Dai-ichi Nuclear Power Plant (NPP) (March 11, 2011) similarly to the accident at the Chernobyl NPP (April 26, 1986) is related to the level 7 of the INES. It is of interest to make an analysis of the radionuclide composition of the fallout following the both accidents. The results of the spectrometric measurements were used in that comparative analysis. Two areas following the Chernobyl accident were considered: (1) the near zone of the fallout - the Belarusian part of the central spot extended up to 60 km around the Chernobyl NPS and (2) the far zone of the fallout--the "Gomel-Mogilev" spot centered 200 km to the north-northeast of the damaged reactor. In the case of Fukushima accident the near zone up to about 60 km considered. The comparative analysis has been done with respect to refractory radionuclides (95Zr, 95Nb, 141Ce, 144Ce), as well as to the intermediate and volatile radionuclides 103Ru, 106Ru, 131I, 134Cs, 137Cs, 140La, 140Ba and the results of such a comparison have been discussed. With respect to exposure to the public the most important radionuclides are 131I and 137Cs. For the both accidents the ratios of 131I/137Cs in the considered soil samples are in the similar ranges: (3-50) for the Chernobyl samples and (5-70) for the Fukushima samples. Similarly to the Chernobyl accident a clear tendency that the ratio of 131I/137Cs in the fallout decreases with the increase of the ground deposition density of 137Cs within the trace related to a radioactive cloud has been identified for the Fukushima accident. It looks like this is a universal tendency for the ratio of 131I/137Cs versus the 137Cs ground deposition density in the fallout along the trace of a radioactive cloud as a result of a heavy accident at the NPP with radionuclides releases into the environment. This tendency is important for an objective reconstruction of 131I fallout based on the results of 137Cs measurements of soil samples carried out at

  6. Radionuclides contamination of fungi after accident on the Chernobyl NPP

    Accumulation of radionuclides by the higher fungi (macromycetes) after the accident on the Chernobyl atomic power plant in 1986 has been studied. Researches were spent in territory of the Chernobyl alienation zone and the Kiev region. Our research has shown that macromycetes accumulate almost all types of radionuclides originating from the accident (131I, 140Ba /140La, 103Ru, 106Ru, 141Ce, 144Ce, 95Nb, 95Zr, 137Cs and 134Cs). They accumulate the long-living 90Sr in much smaller (to 3 - 4 orders) quantities than 137Cs. We have established existence of two stages in accumulation of 137Cs by higher fungi after the accident on the Chernobyl NPP: the first stage resides in the growth of the concentration, the second - in gradual decrease of levels of specific activity of this radionuclide. Despite reduction of 137Cs specific activity level, the content of this radionuclide at testing areas of the 5-km zone around the Chernobyl NPP reaches 1,100,000 Bq/kg of fresh weight in 2013. We investigated dynamics of accumulation of Cs-137 in higher fungi of different ecological groups. One of the major factors that influence levels of accumulation of 137Cs by fungi is their nutritional type (ecological group). Fungi that belong to ecological groups of saprotrophes and xylotrophes accumulate this radionuclide in much smaller quantities than symbio-trophic fungi. As a result of the conducted research it has been established that symbio-trophic fungi store more 137Cs than any other biological objects in forest ecosystems. Among the symbio-trophic fungi species, species showing the highest level of 137Cs contamination vary in different periods of time after the deposition. It is connected with variability of quantities of these radio nuclides accessible for absorption at the depth of localization of the main part of mycelium of each species in a soil profile. Soil contamination by 137Cs is one of the principal abiotic influences on the accumulation of this radionuclide by fungi

  7. Separation of plutonium from uranium and fission products in the zirconium pyrophospate column

    Distribution coefficients were of the following ions were determined in the system zirconium pyrophosphate - aqueous solution HNO3 : Pu3+, Pu4+, PuO22+, UO22+, 234Th2+, 95Zr, 95Nb, 106Ru, 144Ce3+, 90Sr2+, 137Cs+, 59Fe3+ and 59Fe2+. According to the distribution coefficients it can be concluded that the separation of some cations is possible. This was proved by using separation columns. The following successful separations were completed: 90Sr2+ from 90I3+, 90Sr2+ from 90I3+ and 1'37Cs+, UO2+ from 234Th4+, Pu4+ from UO22+, 95Zr, 95Nb, 106Ru, 144Ce3+, 90Sr2+, 137Cs+. Decontamination factors of plutonium from the mentioned cations were determined. It was found that the sorption of Cs+ and Sr2+ is based on ion exchange

  8. Studies on screening of adsorbents for up-take of radioruthenium during vitrification of high-level radioactive waste

    106Ru (T1/2 = 369 days) and 103Ru (T1/2 = 39.3 days) are high specific activity fission products present in different types of nuclear waste generated during the reprocessing of spent nuclear fuel. A major part of it converts into RuO4 vapour during vitrification of high-level waste. Different types of materials like zeolite, perchloric acid treated siliceous brick, stainless steel, polythene, polypropylene, silicone were used as an adsorbent where RuO4 reduced to RuO2 and gets deposited on them which subsequently minimizes the spread of contamination. (author)

  9. A Dosimetry of 106-Ru - 106- Rh Electron-Photon Field with LiF TLD-100 'Microcubes'

    Background and purpose: Ru-Rh eye applicators are used for the radiotherapy of eye malignancies such as melanomas. The producer (BEBIG GmbH) declares ±30% dose uncertainty in the applicator certificate. There is an obvious need to overcome this large imprecision. Our goal is to establish a method that is fast and reliable and which reduces dose uncertainty to below ±10%. Materials and methods: A pleksiglas phantom containing spherical calottes was constructed for this purpose. It allows measurements of surface homogeneity, absolute surface dose rates and depth doses in 2 mm steps, on and off the symmetry axis. Very small, 1x1x1 mm3 thermoluminescent dosimeters (TLD) were chosen as dosimeters. They have to be calibrated adequately for Ru-Rh dosimetry. To do that, 60Co/electron field response ratio of TLDs was investigated and correction factors were established. Doses to the base of optic nerve were considered. Results: The 60Co/electron beam response ratio of TLDs is in agreement with measured and calculated results reported by other authors (for LiF TLD-100 of different dimensions). The measurements of surface dose rate homogeneity show deviations of up to 15% of the mean surface dose rate. However, surface inhomogeneities average out deeper in the phantom. Absolute surface dose rates were found lower than those declared by the producer by 6.2%, averaged over 8 applicators investigated. On and off axis depth doses are highly uniform over angle θ=40 degrees from symmetry axis, up to 6 mm in depth. Therefore, depth doses form spherical isodose surfaces within stated angle, spanning at least 1.47 steradians. Depth dose functions were interpolated to the measured data for practical routine use. Conclusions: The method developed here has overall combined uncertainty <±6%, and therefore reduces significantly uncertainty declared by the producer. Also, it proved to be stable on repeated measurements. (author)

  10. Characterization of Inner Tracker silicon prototype sensors using a 106 Ru-source and a 1083 nm laser system

    Bauer, C; Pugatch, V; Schmelling, M; Schwingenheuer, B; Sievers, P

    2001-01-01

    Silicon strip sensors will be used as technology for the LHCb Inner Tracker. The signal-to-noise ratio (S/N) and the charge division among two neighbouring strips are important sensor parameters. In order to investigate the charge sharing between two adjacent strips, measurements with a 1083 nm laser were performed. Minimum ionizing electrons from a ruthenium-source were used to study the S/N-performance of the silicon sensors. The S/N-measurements were also performed with sensors after irradiation with 24 GeV protons up to a fluence of 1.7*10^14 cm^-2, corresponding to more than 10 years of LHCb Inner Tracker operation in the most irradiated zones.

  11. Measurement of dose rate from 106Ru/106Rh ophthalmic applicators by means of alanine-polymer foils

    Ophthalmic applicators are used in radiotherapy for the treatment of malignant choroidal melanomas. The applicators are positioned on the eye at the base of the tumor for a period of a few days up to 2 weeks. They are commercially available in the form of caps of a spherical radius of 12 or 13 mm. Two or three fixing lugs are used for suturing the applicator to the eye. The applicators are made of silver. The active layer is covered with 0.1 mm silver in the concave direction, and 0.9 mm silver in the convex direction. The β-radiation emitted from the concave side may be used for treatment, to a depth up to 5 mm. Measurements of dose rate by means of ESR/alanine dosimetry and TLD are described. It is concluded that ESR/dosimetry and TLD are well suited for this application. (author)

  12. Development of removal methods of radioactive ruthenium by using the column packed with cell materials

    Ruthenium is an element of various valencies and present in many chemical species of nitro-nitrosyl complexes in nitric acid or in solutions containing nitrates. Since radioactive ruthenium (103Ru, 106Ru) of those chamical species is contained in the wastes occurred on the fuel reprocessing by Purex method and others, it is one of the nuclides which are most difficult to be removed by the conventional methods of the radioactive waste treatments. It was found that this nuclide was effectively removed by passing the waste solution through a column packed with the mixture of powder of anode and cathode materials and depolarizers used in the electric cells. The typical mixtures were zinc-charcoal, zinc.palladium-charcoal, zinc-manganese dioxide.charcoal and zinc-carbon fluoride.charcoal. These column methods showed a surpassing removel efficiency for 106Ru complexes and fisson products. The decontamination factors of radioactive ruthenium were 104 for all kinds of ruthenium complexes and 102 for the species not easily removed by the conventional methods. It was also found that the concentrations of 239Pu, U, 144ce, 155Eu, and 125Sb in the waste could be decreased to that below the limits of detection by the cell material columns. Because 106Ru of chemical species which was difficult to be removed by conventional methods could be efficiently separated from the waste solutions, it was concluded that the columns packed with cell materials are valuable tools in the radioactive waste treatments. (author)

  13. Influence of complex formation on extraction of some fission products by sorption on inorganic sorbents

    Sorption of fission products of radionuclides 137Cs, 89,90Sr, 90,91Y, 86Rb, 133Ba, 95Zr+95Nb, 95Nb, 103,106Ru, 141,144Ce, 115mCd, 113Sn, 125Sb by hydroxides Fe(III), Mn(IV) on the background of 1 mol/l of NaNO3 at the pretense of ions SO42-, C2O42- at a wide ph range (1+14) is studied in present work. Optimal conditions of extraction of each radionuclide by sorption on inorganic sorbents are defined.

  14. 4. Quaterly progress report 1983

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, sevrage water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 58Co, 60Co, 75Se, 103Ru, sup(110m)Ag, 124Sb, 125Sb, 134Cs, 137Cs, 144Ce, 90Sr, 95Nb, 106Ru, 226Ra, 54Mn, U, K and T). This report exposes also the state of surveillance and assistance operations on work sites and, the state of incidents along the three months; a bibliographic selection is also presented

  15. Separation and purification of fission 99Mo from neutron irradiated UAl3 alloy

    A method has been developed for the separation of fission product 99Mo from irradiated uranium aluminum alloy. The method consists of dissolution of the irradiated target in 6 M NaOH, whereby only aluminium along with 99Mo, 131I and 103Ru get into the solution with traces of 95Zr, 95Nb and 132Te, while all other fission products, activation products (239Np) and uranium remain as solid residue. Al(OH)3 precipitation at lower pH (8-9) removed some of the impurities, e.g. 95Zr, 95Nb, 132Te while AgI/AgIO3 precipitation removed almost all the 131I. 103Ru was removed by addition of NaBiO3 and evaporation to dryness. Subsequently 99Mo was purified by precipitation as Mo-α-benzoin oxime which was dissolved in dilute NaOH. This was subjected to organic impurity and trace iodine separation by passing through silver coated activated charcoal. Final purification was carried out by anion exchange separation. 99Mo was obtained with an overall recovery of 80%. Purity of the 99Mo product was found to be in agreement with the US and European pharmacopoeia. (author)

  16. Precipitation separation of molybdeum-99 by α-benzoinoxime in simulated radioactive solution

    This study investigated separation efficiency of molybdenum-99 and removal percentage of the other nuclides with the adding methods of α-benzoinoxime and the dissolution methods of precipitate formed in a simulated radioactive solution composed of eight elements (Mo, I, Ru, Zr, Ce, Cr, Nd, Sr) that was added tracer level radioactive iostopes. Molybdenum-99 could be separated perfectly by α-benzoinoxime regardless of adding method as molybdenum-99 of 100 % was precipitated. Physical treatment like the adding method of α-benzoinoxime didn't affect the precipitation behavior of the other nuclides such 131I, 103Ru, and 95Zr. Precipitation process of molubdeum-99 by α-benzoinoxime was optimal at the batch type adding and dissolution in 0.4 N sodium hydroxide during 20 minutes. At this condition, molybdenum-99 of 97.1% was separated and the decontamination factor of 131I, 103Ru, and 95Zr was 4.8, 45.5 and 27.8, respectively

  17. Distribution of radionuclides among green alga, marine sediments and sea water

    Distributions of radionuclides among green alga (Ulva pertusa), marine sediment and sea water were examined in laboratory experiments with radioisotope tracers to look into the behavior of the radionuclides released into the coastal sea. Marine sands from the five different seashores along the coast of Japan were used. Distributions of 60Co, 95Zr-95Nb and 106Ru-106Rh in these sands were by far the highest of the three components (marine sand, sea water and marine alga). The distributions of the three radionuclides had not so large fluctuations among the five marine sands and were considered to be rather constant. The total amount found in marine sand and green alga was about 90% for the three radionuclides but about 20% for 137Cs on the average in two weeks after the start of the experiments. The activity ratio of marine sediment (radioactivity in 1 g of sediment/radioactivity in 1 ml of sea water) was 4000 for 60co, 30 for 137Cs, 1800 for 95Zr-95Nb and 900 for 106Ru-106Rh. (auth.)

  18. The impact of Chernobyl fallout on Mytilus sp. collected from the French coast

    The spatial distributions of 103Ru, 106Ru, 110mAg, 134Cs and 13/Cs levels in Mytilus sp. samples gathered from the entire French coast subsequent to the explosion of unit 4 at the Soviet nuclear power plant in Chernobyl are largely in line with those based on aerosol and terrestrial measurements. Thus measurements of this bio-indicator found Channel levels to be lower by a factor of 10 than those observed on the eastern section of the Mediterranean coasts. Areas subject to chronic low-level industrial radioactive discharges, such as the Cotentin shoreline and the Rhone estuary, display a slight increase of 106Ru and 137Cs radioactivity levels. In contrast, contamination is especially clear-cut in areas initially free from radionuclides (i.e. surrounding Nice) where biological removal with time has been characterized and modelized, for both ruthenium isotopic forms, and covering 10 and 94 day biological periods. These data, when compared with those found in the literature, indicate that the memorizing time in the event of an accidental release of ruthenium is shorter than for chronic releases. An assessment of public health implications following the ingestion of mussels labelled solely by Chernobyl fallout is of the same order of magnitude as that of mussels gathered in an area exposed to Rhone industry, i.e. 10-5 of the annual dose limit to the whole body for the public at large

  19. 1975 progress report: Idaho National Engineering Laboratory site radioecology--ecology programs

    Results are reported from measurements of the content of various radionuclides in the tissues of wild animals on or near the Idaho National Engineering Laboratory sampled during 1975. Tissue samples from antelope, waterfowl, rodents, rabbits, and doves were analyzed for 13 radionuclides, including 134Cs, 137Cs, 95Zr, 95Nb, 103Ru, 238Pu, 239Pu, 90Sr, 131I, and 60Co which were responsible for the largest amounts of radioactivity. Measurements were also made of the content of 238Pu, 239Pu, and 241Am in soil samples and the radioactivity in tumbling weeds at the radioactive waste management site. Data are included from studies on the ecology of the pygmy rabbit, Salvilagus idahoensis, amphibians, reptiles, birds of prey, rodents, and coyotes, and vegetation in relation to land use at the site. Seasonal variations in the deposition and retention of 141Ce and 134Cs on sagebrush and bottlebrush grass were compared

  20. Measurement by γ spectrometry of specific activities of radioisotopes present in vegetal ashes. Study of variations of the ambient radioactivity level in the Grenoble transverse valley from March 1966 to August 1968

    The first part of this report addresses the dosimetry of γ emitting radio-elements which are present in vegetal samples. The dose measurements were performed by spectrometry and results were processed by using a least square method. The second part reports works performed in the Grenoble transverse valley by using the same techniques. Radioactivity fluctuations of various radio-elements (40K, 54Mn, 95Zr + 95Nb, 103Ru + 106Rh, 137Cs, 137Ba, 140Ba + 140La, 144Ce + 144Pr) in various vegetal species, in water and in sediments have been monitored in seven points of the Isere river banks, upstream and downstream the city of Grenoble, from March 1966 to August 1968. Fluctuations observed for each radio-element are explained by comparison with physiologic, hydrologic and atmospheric climate conditions. The principles of a systematic control of a site for the detection of possible radioactive pollutions are then defined

  1. Studies on inorganic exchangers - manganese dioxide

    As a part of investigation of separation processes for long lived fission products from fuel reprocessing solution, manganese dioxide has been studied as an ion exchanger for cerium using 137Cs, 106Ru, 141Ce, sup(85,89)Sr, 95Zr and 95Nb as tracers. For different concentrations of HNO3, distribution ratios and breakthrough capacities were determined. Cerium was eluted by manganese sulphate and nitric acid. Results show that : (1) at all acidities cerium is adsorbed with almost no uptake of other rare earths, sodium, uranium and plutonium, (2) Ce (IV) gives better adsorption than Ce(III), (3) a combination of manganese sulphate (1 mg/ml) and 3M nitric acid elutes 99% cerium in 5-6 column volumes and (4) as for effect of absorption-elution cycles on MnO2 column, initially there is a decrease in capacity of cerium uptake but thereafter the capacity remains constant. (M.G.B.)

  2. Measurements of fission product concentrations in surface air at Bombay, India, during the period 1975-1981

    Measurements on airborne fallout radioactivity for the period 1975 up to the middle of 1981 are given. Normally, these measurements are confined to Bombay, but after nuclear tests, some of the other stations where these measurements were carried out in previous years are operated for some time to study the levels of fresh activity. The levels of the long-lived fission products 144Ce, 106Ru and 137Cs, and the short-lived fission products 95Zr and 140Ba, were measured, whenever they could be detected following nuclear tests, and tabulated. The data indicate that the activity varies by large factors from tests of similar yield, depending on the meteorological and other conditions. It was determined that the travel time for the Chinese test debris from Lop Nor, China to the West-coast of India is 14 to 16 days

  3. Preparative electrophoresis of industrial fission product solutions

    The aim of this work is to contribute to the development of the continuous electrophoresis technique while studying its application in the preparative electrophoresis of industrial fission product solutions. The apparatus described is original. It was built for the purposes of the investigation and proved very reliable in operation. The experimental conditions necessary to maintain and supervise the apparatus in a state of equilibrium are examined in detail; their stability is an important factor, indispensable to the correct performance of an experiment. By subjecting an industrial solution of fission products to preparative electrophoresis it is possible, according to the experimental conditions, to prepare carrier-free radioelements of radiochemical purity (from 5 to 7 radioelements): 137Cs, 90Sr, 141+144Ce, 91Y, 95Nb, 95Zr, 103+106Ru. (author)

  4. Transuranic elements and strontium-90 in samples from forests in Poland

    Enhanced levels of non-volatile nuclides;141,144Ce, 95Zr, 95Nb, 103,106Ru, 238,239,240,241Pu, 241Am, 242,243,244Cm and 154,155Eu were observed in the samples from north-western Poland. This was considered to be a result of finding in sample ''hot particle''. Investigations conducted in the Institute of Nuclear Physics allowed us to conclude that on this area a non typical isotopic composition of Chernobyl fallout was very common. The enhanced activities (up to 100% above the global fallout value) of 90Sr, 238,239,240Pu and 241Am were observed. The presence of 243,244Cm and 154,155Eu was confirmed. It seems that the quasi-continuous fallout of huge numbers of small 'hot particles' occurred there from the high altitude radioactive cloud, which moved toward Scandinavia on 26-th of April, 1986. (author)

  5. Radiation accidents on human in the nuclear installations and their medical emergency procedures, (1)

    Present nuclear installations are one of the safest installations among industrial facilities, being equipped with various safety instruments. Since X-ray was discovered in 1895, however, many radiation injuries of various degrees and kinds occurred. Among dangerous nuclides often observed as radioactivity pollutions in nuclear installations, the exposure to β-ray such as 90Sr, 106Ru, 95Zr, 131I, 144Ce, etc, is considered to be serious problems. When they affect wounds or are inhaled into lungs, only symptomatic treatment is practicable at present, and usually nothing can be depended upon, but spontaneous eliminating ability. As the mass inhalation of α nuclides, especially transuranium nuclides, is quite dangerous, the treatment by lung-irrigation now under development is most effective as the emergency treatment. When trans-uranium nuclides were accidentally observed from wounds, they should be eliminated by the injection of chelating agent. (Kobatake, H.)

  6. Concentration and depuration of some radionuclides present in a chronically exposed population of mussels (Mytilus edulis)

    Factors are described which affect the concentration (p Ci g-1 dry wt) and loss of 241 Am, 239+240Pu, 238Pu, 144Ce, 137Cs, 134Cs, 106Ru, 95Zr and 95Nb in an exposed population of mussels Mytilus edulis L. from Ravenglass on the Esk estuary, Cumbria, UK which receives radioeffluents from the British Nuclear Fuels Ltd. (BNFL) plant at Sellafield, some 10 km to the north. Tidal position and mussel body size have a negligible influence on the concentration of 241Am, 137Cs and 106Ru in the total soft tissue, but variation in soft tissue weight throughout the year has a considerable influence on the apparent concentration and depuration times of these radionuclides. Apart from the clearance (tsub(1/2) biol, 1 to 3 h) of sediment-associated activity from the digestive tract, the depuration rate profiles follow a single component clearance curve with a biological half-life in excess of 200 d for 241 Am, 239+240Pu, 238Pu and 144Ce, and of 40 d for 137Cs. The clearance of 106Ru is more complex and consists of a 3 component depuration profile with biological half-lives of 6 h, 12 d and 260 d. The depuration profiles presented in this work are for chronically ingested isotopes under natural conditions; acute exposure will most likely result in different profiles, especially those derived from laboratory spiking experiments. Isotope ratio data support the hypothesis that the main route of entry into the mussel for the majority of the radionuclides studied is from the water. (orig./WL)

  7. Transfer of radionuclides from sediments to organisms

    The following items were investigated: 1) the transfer of radionuclides (60Co, 95Zr-95Nb, 106Ru-106Rh, 37Cs, and sup(115m)Cd) from sea water to marine sediments, 2) the transfer of the radionuclides from polluted sediments to the organisms, and 3) the transfer of the radionuclides from polluted sea water to the organisms. Earth of the sea bottom, which has a diameter of 0.1 to 0.5 mm phi, was collected as sediment, and Nereis japonica and green alga were used as the organisms. 1) Activity ratio for sediment, which fits for concentration factor the organism, was at its maximum in 60Co, and it decreased in order of 95Zn-95Nb, 106Ru-106Rh, and 137Cs. The distribution of sup(115m)Cd in 500 ml of sea water, 30 g of marine sediment, and 3 g of green alga was 60, 32, and 8%, respectively. Activity ratio of sup(115m)Cd for marine sediment was 9, and concentration factor of sup(115m)Cd for green alga was 21. 2) With respect to radionuclides transfer from sediments to the organism, the level of radionuclides in Nereis japonica reached the equilibrium in about one week. Radioactivity concentration in polluted sediments at the beginning of this experiment and that in Nereis japonica 11 days later were expressed by Acpm/g and Bcpm/g, respectively, and B/A (transfer ratio) was calculated. Transfer ratio of 137Cs, 60Co, 95Zr-95Nb, and 106Ru-106Rh were 0.175, 0.050, 0.009, and 0.06, respectively. The transfer ratio of 137Cs was small when 137Cs transferred from sea water to marine sediments, but the ratio of 137Cs dissolved again into sea water in spite of being absorbed into sediments was larger than that of other nuclides. The transfer ratio of sup(115m)Cd was 0.12. 3) The transfer of nuclides to Nereis japonica was influenced strongly by sea water more than sediments. (Tsunoda, M.)

  8. Reconstruction of radionuclide releases from the Hanford Site, 1944-1972

    Historic releases of key radionuclides were estimated as a first step in determining the radiation doses that results from Hanford Site operations. The Hanford Site was built in southcentral Washington State during World War II to provide plutonium for the U.S. nuclear weapons program. As part of the Hanford Environmental Dose Reconstruction (HERD) Project, releases to the Columbia River of 24Na, 32P, 46Sc, 51Cr, 56Mn, 65Zn, 72Ga, 76As, 90Y, 131I, 239Np, and nonvolatile gross beta activity from operation of eight Hanford single-pass production reactors were estimated. Releases of 90Sr, 103Ru, 106Ru, 131I, 144Ce, and 239Pu to the atmosphere from operation of chemical separation facilities were also estimated. These radionuclides and the atmospheric and Columbia River pathways were selected for study because scoping studies showed them to be the largest contributors to dose from Hanford operations. The highest doses resulted from releases to the atmosphere of 131I from chemical separations plants in the pre-1950 period. Prior to 1950, the technology for limiting iodine releases had not been developed. Hence, a very detailed reconstruction of the hourly 131I release history was achieved for 1944-1949 using Monte Carlo methods. Atmospheric releases of the other radionuclides were estimated on a monthly basis for 1944-1972 using deterministic calculations. Monthly releases to the Columbia River for 1944-1971 were based on Monte Carlo methods

  9. Monte Carlo modeling of beta-radiometer device used to measure milk contaminated as a result of the Chernobyl accident

    This paper presents results of Monte Carlo modeling of the beta-radiometer device with Geiger-Mueller detector used in Belarus and Russia to measure the radioactive contamination of milk after the Chernobyl accident. This type of detector, which is not energy selective, measured the total beta-activity of the radionuclide mix. A mathematical model of the beta-radiometer device, namely DP-100, was developed, and the calibration factors for the different radionuclides that might contribute to the milk contamination were calculated. The estimated calibration factors for 131I, 137Cs, 134Cs, 90Sr, 144Ce, and 106Ru reasonably agree with calibration factors determined experimentally. The calculated calibration factors for 132Te, 132I, 133I, 136Cs, 89Sr, 103Ru, 140Ba, 140La, and 141Ce had not been previously determined experimentally. The obtained results allow to derive the activity of specific radionuclides, in particular 131I, from the results of the total beta-activity measurements in milk. Results of this study are important for the purposes of retrospective dosimetry that uses measurements of radioactivity in environmental samples performed with beta-radiometer devices.

  10. Assessment of Cesium, Iodine, Strontium and Ruthenium isotopes behaviour in urban areas, after contamination from accidental release

    The exposures of urban populations to the radiation derived from the deposition, after accidental atmospheric releases, of 137 Cs, 134 Cs, 129 I, 131 I, 133 I, 89 Sr, 103 Ru and 106 Ru were assessed, using the integrated system for the evaluation of environmental radiological impact in emergency situations, developed by the Instituto de Radioprotecao e Dosimetria (IRD)/Comissao Nacional de Energia Nuclear (CNEN). These radionuclide are fission products likely to be emitted in the occurrence of severe nuclear reactor accidents. Their environmental behaviour in urban areas, due to their deposition in soil, in urban surfaces and in vegetable-garden food products, such as leafy and non-leafy vegetables, were analyzed, and dose assessments at the short, medium and long terms were performed, with an without the application of protective measures for reduction of doses. Simulations of unitary initial deposition for each radionuclide and of two different potential accidents involving water reactors (PWR), with different source terms and distinct deposition for each radionuclide, were performed. Results were analyzed on the basis of relative relevance of radionuclides and pathways for the exposure of members of the public, as a function of age and time after the release. It was also performed an assessment of the effectiveness of protective measures as a function of the moment of their implementation. (author)

  11. Radioactive decontamination methods and their effectiveness as a function of terrain

    A large area of rugged terrain on the Nevada Test Site was contaminated following a spill of radioactively contaminated drilling mud. The contamination was shown to consist of 103Ru and 106[Ru-Rh] with total estimated activity at release time of 38 and 6 Ci, respectively. Several decontamination methods were used and their effectiveness assessed by determining the fraction of radioactivity remaining (FR) following each. In flat areas, the front end loader was by far the most efficient method, removing large quantities of dirt in relatively short periods of time. FRs of 10-22 were achieved. In canyon areas, flushing with water was most effective on rocky surfaces with FRs of 10-3, while shoveling and bagging in evaporated mud collection ponds worked well and resulted in FRs of 10-2. The FR in rocky cracks was about 10-1 following flushing with water. In Locations where radioactive mud/water had not penetrated the ground surface to more than 1-2in., such as fine grain, flat compact dirt, vacuuming was very effective achieving FRs of 10-3. However, unless the contaminated area was very small (e.g. dropping from front end loading operations), vacuuming was too slow to be of practical value. Under the supervision of experienced radiation monitors, the radioactive mud spill area was safely cleaned up using, for the most part, standard earth moving equipment and personnel untrained in decontamination procedures. (author)

  12. The application of an environmental radiological impact integrated evaluation system in emergency situations for the radiological exposure evaluation of urban areas population

    An integrated computational system for environmental radiological impact assessment in emergency situations has been developed at IRD/CNEN. This system has been used, in this paper, to assess the exposure to members of the public, in a urban area, after the release to the environment of radionuclides that may be relevant after a severe nuclear accident with a PWR reactor. Doses were calculated, for the short and long terms, for the radionuclides 137 Cs, 134 Cs, 129 I, 131 I, 133 I, 89 Sr, 90 Sr, 103 Ru and 106 Ru. The study also simulates the application of some protective measures aiming the estimate of dose reduction, taking into account the moment of their implementation. The starting point for the simulation is the activity deposited on the ground and on urban surfaces and includes the ingestion of home grown products, such as green vegetables, legumes and chicken. This paper is part of a program to derive tools for planning of emergency attendance, taking into account the present knowledge about the fate of radionuclides delivered to the environment, based on estimates of exposure, for the protection of population in tropical climate environments. (author)

  13. Effect of sterilization by gamma-irradiation on the sorption of 137Cs, 85Sr, 139Ce, 57Co, 109Cd, 65Zn, 103Ru, 95mTc and 131I by soils

    Six soils, two Sphagnum peat samples and a clay mineral were irradiated with 40 and 80 kGy (4 and 8 Mrad) from a 60Co source. As a result the microbial biomass, determined separately for each sample, decreased considerably. Depending on the radionuclide, the sorption, as characterised by the distribution coefficient, decreased, increased or remained unchanged. The effect of the irradiation on the sorption of the radionuclides depended, in general, also on the type of the sample, especially whether well humified soils, (e.g. crop soils), poorly humified samples (Sphagnum peat, O-horizon from woodland), or a clay mineral was employed. The data reveal that irradiation produces, besides sterilization, also other effects in soils, which can change their sorption properties. (orig.)

  14. Environmental gamma radiation and fallout measurements in Finland, 1986-87

    Results from a survey of environmental gamma radiation levels in Finland after the Chernobyl accident 1986 are presented. The measurements were made in 1986-87 by means of sensitive Geiger-counters and a gamma-spectrometer placed in cars. The results show the level of external radiation caused by the cesium fallout on the first of October 1987. The fallout pattern of 137Cs as well as of 95Zr and 103Ru are also presented. In the center of Southern Finland there are wide areas with exposure levels exceeding 0.03 μSv h-1, areas exceeding 0.10 μSv h-1 being very rare. The surface area weighted mean dose rate for the 461 municipalities in Finland was 0.027 μSv h-1 (range 0-0.19 μSv h-1). The population weighted mean dose rate was 0.037 μSv h-1. The corresponding estimated surface activity of 137Cs was 10.7 kBq m-2. The passage of the Chernobyl plume over Finland in 1986 led to various fallout patterns for different radionuclides. The deposition of the non-volatile nuclides, 95Zr and 141Ce, is closely related to the passage of the hot particle dust formed at the initial explosion in the reactor at 01.23 LT on 26 April. This cloud passed over Finland between the morning and the night of 27 April. The deposition of volatile fission products such as 131I, 132Te, 134Cs and 137Cs in Finland was caused by releases from the burning reactor after the initial explosion. The radioactive plume spread over Southern and Central Finland between Sunday 27 April and Tuesday 29 April. On 30 April and finally on 1 May a could northerly airstream spread into the whole of Finland purifying the atmosphere. The volatile nuclides were mainly deposited by intermittent rain on 28-30 April. The deposition pattern of 103Ru is a combination of the fallou patterns due to the initial explosion and the reactor burn, as well as the wet deposition occurring on 10-12 May caused by the releases from the burning reactor in early May

  15. Health risks from radionuclides released into the Clinch River

    The purpose of this work is to estimate off-site radiation doses and health risks (with uncertainties) associated with the release of radionuclides from the X-10 site. Following an initial screening analysis, the exposure pathways of interest included fish ingestion, drinking water ingestion, the ingestion of milk and meat, and external exposure from shoreline sediment. Four representative locations along the Clinch River, from the White Oak Creek Embayment to the city of Kingston, were chosen. The demography of the lower Clinch River supplied information dealing with land use that aided in the determination of sites on which to focus efforts. The locations that proved to be the most significant included Jones Island at Clinch River Mile (CRM) 20.5, Grassy Creek and K-25 (CRM 14), Kingston Steam Plant (CRM 3.5), and the city of Kingston (CRM 0). These areas of interest have historically been and are still primarily agricultural and residential areas. Reference individuals were determined with respect to the pathways involved. The primary radionuclides of interest released from the X-10 facility into the Clinch River via White Oak Creek were identified in the initial screening analysis as 137Cs, 90Sr, 60Co, 106Ru, 144Ce, 131I, 95Zr, and 95Nb. Of these radionuclides, 137Cs, 60Co, 106Ru, 90Sr, 144Ce, 95Zr, and 95Nb were evaluated for their contribution to the external exposure pathway. This study utilized an object-oriented modeling software package that provides an alternative to the spreadsheet, providing graphical influence diagrams to show qualitative structure of models, hierarchical models to organize complicated models into manageable modules, and intelligent arrays with the power to scale up simple models to handle large problems. The doses and risks estimated in this study are not significant enough to cause a detectable increase in health effects in the population. In most cases, the organ does are well below the limits of epidemiological detection (1 to 30

  16. Study on separation of 137Cs from 235U fission process waste - utilization of silica gel-supported ferrocyanide complex salt for 137Cs picking

    In connection with the potential domestic demand especially in the fields of industry and nuclear medicine, the separation of 137Cs from 235U fission process waste is to be of interest although its economic value could be a polemic. A preliminary study on the separation of 137Cs from the 235U fission process waste generated in the production of 99Mo in P.T. BATAN Teknologi, Serpong, was performed through experiments on 137Cs picking from sample solution of the radioactive fission waste (RFW). The presented study is aimed to gain experimental data supporting utilization of the matrix of silica gel-supported ferrocyanide complex salt for the separation of 137Cs from RFW. Subsequent step would be the recovery and purification of 137Cs as part of production technology of 137Cs. The RFW sample was batch-treated with the matrix of silica gel-supported ferrocyanide complex salt which was synthesized from silica gel, potassium hexacyanoferrate(II) and copper(II) chloride. The binding of radioisotopes in RFW on the matrix was observed by γ-spectrometry of the RFW solution before and during the treatment. The results showed that approximately 85% of 137Cs could be picked from the RFW sample into the matrix. Less amount of 95Zr and 95Nb was bound into the matrix. 103Ru was slightly bound into the matrix whereas 141/144Ce and 129mTe were not. It was observed that by using 0.2 and 0.4 g of matrix for 10 ml of RFW, the amount of matrix influenced the binding quantity of 95Zr and 95Nb but not that of 137Cs. (author)

  17. Preliminary studies on adsorption of Ruthenium on carbon nanotubes

    Commercial availability of carbon nanotubes (CNTs) in different forms has enhanced the research on its ability to treat effluents. Due to porous and hollow structures, large surface area, low density, high mechanical, thermal and chemical stabilities, they are being used as a potential adsorbent for removing a wide spectrum of both organic compounds and inorganic ions. Examples include: organic compounds such as dioxin, resorcinol and other phenolic derivatives, pesticides and metallic ions such as lead, copper etc. Recently they have also been applied for recovery of radionuclides such as thorium, europium, americium and plutonium from aqueous solutions. They have also been functionalised with different groups such as diglycolamide for effective adsorption of uranium. In continuation of the study, attempt has been made on the adsorption of Ruthenium on pristine carbon nanotubes. Ruthenium (Ru), a rare transition metal of platinum group elements, is typically present in common terrestrial rocks at ng g-1 level. It is also produced as a fission product in nuclear reactors. It has seven naturally occurring isotopes and thirty four radioactive isotopes. Of these radioactive isotopes, the most stable radioisotopes viz, 97Ru (t1/2=3 days), 103Ru (t1/2=40 days) and 106Ru (t1/2=386 days) are important from the environmental point of view. 106Ru, a soft beta emitter (Eβ max = 39 keV) is used for treatment of eye cancer. The highly volatile nature of Ruthenium as RuO4 (B.Pt.313 K) is an important aspect to be considered in spent nuclear fuel reprocessing facilities. Traces of these radionuclides may remain in waste solutions generated in reprocessing laboratories. Among the methods available for separation of Ruthenium, adsorption plays an important role since it eliminates the need for huge sludge handling process. A well designed sorption process with high efficiency results in a high-quality effluent after treatment which can be recycled or safely disposed. Studies on

  18. Solvent extraction study on the separation of molybdenum-99 and uranium in sulfuric acid solution by tri-n-octylamine in kerosene

    The basic extraction study on the separation of fission product molybdenum-99 and uranium in sulfuric acid solution by tri-n-octylamine (TOA) in kerosene has been investigated. The equilibration time and the effect of temperature, concentration of extractant, uranium and sulfuric acid concentration on this extraction system were examined. The optimum conditions for the coextraction of molybdenum-99 and uranium have been obtained with the overall recovery of 90% for 99Mo and greater than 99% for uranium. Based on the complex stability difference between UO2(VI) and MoO2 (VI) with TOA, uranium in the organic phase can be back-extracted by proper chloride concentration. On the other hand, molybdenum-99 can be stripped from the organic phase by sodium carbonate or ammonium hydroxide solution. Decontamination factors of some major fission products such as 95Zr, 95Nb, 103Ru, 132Te, 141Ce and 131I in the separation process were also examined in this report. 11 references, 8 figures, 2 tables

  19. Interim environmental monitoring report for the Nevada test site, first quarter 1981

    During the first calendar quarter of 1981, no radioactivity from the nuclear tests conducted at the Nevada Test Site was measured offsite by the US Environmental Protection Agency's Environmental Monitoring Systems Laboratory. Low concentrations of 95Zr, 95Nb, 103Ru, and 141Ce attributed to the People's Republic of China nuclear test of October 15, 1980, were detected in air samples throughout the Air Surveillance Network. The maximum concentrations of these radionuclides were less than 0.1 percent of the Concentration Guides. The dosimeters of fixed station at Complex I (Coal Valley) indicated an exposure of 1.6 mR, and the dosimeters of two offsite residents, one living at Glendale, Nev., and the other near Complex I, (Coal Valley) appeared to have net exposures of 3.1 mR and 3.2 mR, respectively; however, further evaluation revealed that the net exposures were not due to an exposure from NTS operations, but may be a statistical anomaly related to an unusually low variation in the environmental background exposure rate. Further investigation is in progress

  20. Inorganic oxides as alternative in the separation of non fissioned residual uranium

    The Al2O3, SiO2 and SnO2 as well as vegetable carbon have been studied for its possible use as sorbent in the concentration and separation of non fissioned residual uranium of some fission products such as: 141 Ce, 134 Cs, 125 Sb, 103 Ru, 95 Zr, 95 Nb of alkaline aqueous systems. The separation efficiency has been evaluated using natural uranium and radionuclides in static and dynamic processes, through liquid scintillation and gamma spectrometry. Therefore Al2O3, SiO2, SnO2 and carbon were pre-treated thermic and chemically and characterized through the technique of Nitrogen absorption analysis, X-ray diffraction and IR spectroscopy. By means of the p H determination and the aqueous system potential the present hydrolysis products were determined. The inorganic oxides show structural and surface changes due to the treatment. The adsorption process is realized by different mechanism depending of the sorbent. The results show that the retention capacity is a dependence of the oxides pre-treatment and of the hydrolysis products in the aqueous system, as well as of the experimental conditions. Not in this way for carbon in which the results show the treatment and the experimental conditions significantly have not influence in its adsorption capacity. (Author)

  1. Results of gamma-scanning measurements on Windscale AGR stringer Mk4/S46 and implications for civil AGRs

    Axial rating distributions in a WAGR fuel stringer were determined retrospectively using the measured gamma count rates from five fission products after cooling times of 70 to 100 days. The results from all five isotopes were in general agreement with predictions using the Windscale code CAMPARI, but revealed mutually consistent discrepancies with prediction near the top axial reflector. The measurements were conducted to test the gamma-scanning apparatus and methods of analysis in preparation for the post irradiation examination of CAGR fuel. It was concluded from the results that accuracies of +- 2% to +- 3% (1 sigma) should be attainable in the determination of relative ratings using the measured gamma count rates from the isotopes 140Ba, 103Ru, 95Zr, 144Ce and 137Cs. The accuracies will, however, depend on the type of fuel being examined and the fission product used. In particular, care must be taken when using 137Cs since it may migrate from its point of formation, and cooling times of less than 110 days must be achieved to measure relative concentrations of the short-lived isotope 140Ba with sufficient accuracy. (author)

  2. Inorganic oxides as alternative in the separation of non fissioned residual uranium; Oxidos inorganicos como alternativa en la separacion del uranio residual no fisionado

    Baca G, A

    1997-07-01

    The Al{sub 2}O{sub 3}, SiO{sub 2} and SnO{sub 2} as well as vegetable carbon have been studied for its possible use as sorbent in the concentration and separation of non fissioned residual uranium of some fission products such as: {sup 141} Ce, {sup 134} Cs, {sup 125} Sb, {sup 103} Ru, {sup 95} Zr, {sup 95} Nb of alkaline aqueous systems. The separation efficiency has been evaluated using natural uranium and radionuclides in static and dynamic processes, through liquid scintillation and gamma spectrometry. Therefore Al{sub 2}O{sub 3}, SiO{sub 2}, SnO{sub 2} and carbon were pre-treated thermic and chemically and characterized through the technique of Nitrogen absorption analysis, X-ray diffraction and IR spectroscopy. By means of the p H determination and the aqueous system potential the present hydrolysis products were determined. The inorganic oxides show structural and surface changes due to the treatment. The adsorption process is realized by different mechanism depending of the sorbent. The results show that the retention capacity is a dependence of the oxides pre-treatment and of the hydrolysis products in the aqueous system, as well as of the experimental conditions. Not in this way for carbon in which the results show the treatment and the experimental conditions significantly have not influence in its adsorption capacity. (Author)

  3. Position of sediments in transfer of radionuclides released into coastal sea to human beings

    A great portion of radionuclides released into coastal sea is adsorbed into marine sediments and the adsorbed radionuclides causes the radioactive contamination of marine organisms and then transferred to human beings who consume them. In order to make a quantitative evaluation of this route, the transfer of 9 kinds of radionuclides from sediments to benthic organisms such as algae, bivalve and worm was observed. Then it was compared with the radioactivity in these organisms from radioactively contaminated sea water (Concentration factor). It was observed that the influence of sea water was larger than that of sediments as it was 104 times larger for 54Mn, 103 times larger for 59Fe, 60Co, 95Zr-95Nb, 106Ru-106Rh and 144Ce-144Pr, 102 times larger for 65Zn and 10-102 times larger for 115mCd and 137Cs. Consequently, sea water can be considered as the main route and sediments as the secondary in the case of quantitative comparison of the effect on the accumulation of radionuclides by marine organisms and in the transfer of radionuclides to human beings

  4. Chromatographic isolation of 144Ce and 144Pr from the wastes of irradiated uranium treatment

    A two-step chromatographic technique was elaborated to isolate 144Ce, 144Pr from a solution of uranium fission products in 6M HNO3. The oxidation to Ce(III) by bromate and selective adsorption of 144Ce(IV) on anion exchange column were used to concentrate and purify 144Ce. Some impurities of uranium, 95Zr, 95Nb, 106Ru remain in 144Ce solution after the first step of its isolation. The final purification is achieved by passing the 6M HNO3 solution of 144Ce(IV) through the HDEHP-coated teflon column. The decontamination factors of 144Ce from main fission products are given. 7.2 mCi of (144Ce+144Pr) are recovered from each gram of irradiated uranium trioxide with the yield greater than 99%. An improvement of known generator was carried out to elute a purer 144Pr from maternal 144Ce(IV) adsorbed on the anion exchange column. (author)

  5. Radioactivity in food crops

    Drury, J.S.; Baldauf, M.F.; Daniel, E.W.; Fore, C.S.; Uziel, M.S.

    1983-05-01

    Published levels of radioactivity in food crops from 21 countries and 4 island chains of Oceania are listed. The tabulation includes more than 3000 examples of 100 different crops. Data are arranged alphabetically by food crop and geographical origin. The sampling date, nuclide measured, mean radioactivity, range of radioactivities, sample basis, number of samples analyzed, and bibliographic citation are given for each entry, when available. Analyses were reported most frequently for /sup 137/Cs, /sup 40/K, /sup 90/Sr, /sup 226/Ra, /sup 228/Ra, plutonium, uranium, total alpha, and total beta, but a few authors also reported data for /sup 241/Am, /sup 7/Be, /sup 60/Co, /sup 55/Fe, /sup 3/H, /sup 131/I, /sup 54/Mn, /sup 95/Nb, /sup 210/Pb, /sup 210/Po, /sup 106/Ru, /sup 125/Sb, /sup 228/Th, /sup 232/Th, and /sup 95/Zr. Based on the reported data it appears that radioactivity from alpha emitters in food crops is usually low, on the order of 0.1 Bq.g/sup -1/ (wet weight) or less. Reported values of beta radiation in a given crop generally appear to be several orders of magnitude greater than those of alpha emitters. The most striking aspect of the data is the great range of radioactivity reported for a given nuclide in similar food crops with different geographical origins.

  6. Radioactivity in food crops

    Published levels of radioactivity in food crops from 21 countries and 4 island chains of Oceania are listed. The tabulation includes more than 3000 examples of 100 different crops. Data are arranged alphabetically by food crop and geographical origin. The sampling date, nuclide measured, mean radioactivity, range of radioactivities, sample basis, number of samples analyzed, and bibliographic citation are given for each entry, when available. Analyses were reported most frequently for 137Cs, 40K, 90Sr, 226Ra, 228Ra, plutonium, uranium, total alpha, and total beta, but a few authors also reported data for 241Am, 7Be, 60Co, 55Fe, 3H, 131I, 54Mn, 95Nb, 210Pb, 210Po, 106Ru, 125Sb, 228Th, 232Th, and 95Zr. Based on the reported data it appears that radioactivity from alpha emitters in food crops is usually low, on the order of 0.1 Bq.g-1 (wet weight) or less. Reported values of beta radiation in a given crop generally appear to be several orders of magnitude greater than those of alpha emitters. The most striking aspect of the data is the great range of radioactivity reported for a given nuclide in similar food crops with different geographical origins

  7. A study of gamma-emitting radionuclides present into the sediments and algae of the ''Baie de l'Orne'' (Central Normandy Coast) collecting during the years 1980 - 1982

    The present status of some environmental effects of existing sources of gamma-emitting radionuclides, along the Central part of the Normandy Coast (Calvados shores and river Orne mouth) was determined. A systematic study was made on the behaviour of the marine sediments and brown alga 'Laminaria digitata' with regard to their properties as indicators of radioactive contamination. Marine sediments were collected into the river Orne at fixed locations and into the sea in and around the estuary from 1977 to 1982. Algae samples were picked up along the coast from 1980 to 1982, mostly on the western part of the Orne estuary. Dosimetry techniques employed have produced accurate and reliable results, despite the very low levels of activity involved. Gamma-emitting radionuclides present in the environment (chiefly 106Ru, 137Cs, 95Zr) were identified and measured. Their behaviour into the marine sediments and into Laminaria Digitata were determined. This study provides evidence on the presence of numerous gamma-emitting radionuclides into the marine environment of the Central part of the Normandy coast, but, altogether with very low levels of radioactivity

  8. Specific activity and activity ratios of radionuclides in soil collected about 20 km from the Fukushima Daiichi Nuclear Power Plant: Radionuclide release to the south and southwest

    Tagami, Keiko, E-mail: k_tagami@nirs.go.jp; Uchida, Shigeo; Uchihori, Yukio; Ishii, Nobuyoshi; Kitamura, Hisashi; Shirakawa, Yoshiyuki

    2011-10-15

    Soil samples at different depths (0-2, 5-7 and 10-12 cm) were collected from J Village, about 20 km south of Fukushima Daiichi Nuclear Power Plant (FNPP) to determine their radionuclide specific activities and activity ratios. The concentrations and activity ratios of {sup 131}I, {sup 134,} {sup 136,} {sup 137}Cs and {sup 129m}Te were obtained, but only trace amounts of {sup 95}Nb, {sup 110m}Ag and {sup 140}La were detected which were too low to provide accurate concentrations. Radionuclides such as {sup 95}Zr, {sup 103,} {sup 106}Ru and {sup 140}Ba that were found in Chernobyl fallout, were not found in these soil samples. This suggests that noble gasses and volatile radionuclides predominated in the releases from FNPP to the terrestrial environment. The average activity ratios of {sup 131}I/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 136}Cs/{sup 137}Cs and {sup 129m}Te/{sup 137}Cs were 55, 0.90, 0.22 and 4.0 (corrected to March 11, 2011) in the 0-2 cm soil samples of April 20 and 28, 2011.

  9. Specific activity and activity ratios of radionuclides in soil collected about 20 km from the Fukushima Daiichi Nuclear Power Plant: Radionuclide release to the south and southwest

    Soil samples at different depths (0-2, 5-7 and 10-12 cm) were collected from J Village, about 20 km south of Fukushima Daiichi Nuclear Power Plant (FNPP) to determine their radionuclide specific activities and activity ratios. The concentrations and activity ratios of 131I, 134,136,137Cs and 129mTe were obtained, but only trace amounts of 95Nb, 110mAg and 140La were detected which were too low to provide accurate concentrations. Radionuclides such as 95Zr, 103,106Ru and 140Ba that were found in Chernobyl fallout, were not found in these soil samples. This suggests that noble gasses and volatile radionuclides predominated in the releases from FNPP to the terrestrial environment. The average activity ratios of 131I/137Cs, 134Cs/137Cs, 136Cs/137Cs and 129mTe/137Cs were 55, 0.90, 0.22 and 4.0 (corrected to March 11, 2011) in the 0-2 cm soil samples of April 20 and 28, 2011.

  10. Kinetics of ruthenium in humans

    The fission products 103Ru and 106Ru may represent a radiological hazard for the population in case of their release in the environment and transfer to the food chain; considerable amounts of these radionuclides were indeed found in the fall-out of the Chernobyl accident and also after nuclear weapon tests. As for many other radionuclides, the biokinetic model for ruthenium currently recommended by ICRP is mainly based on animal data. In order to obtain valuable information directly in humans, a technique based on stable isotopes administration and analytical methods such as activation analysis with charged particles and inductively coupled mass spectrometry has been purposely developed. A total of six investigations were conducted on two healthy volunteers by administration of given amounts of isotopically enriched ruthenium. Blood samples were withdrawn at fixed intervals in the first 6 hours after administration, and the total renal excretion in the first two days collected. Tracer concentration in the biological samples were determined using the analytical methods introduced above. The results show that the kinetics in plasma and the urinary pattern present several deviation from the behaviour described by the current ICRP model. The intestinal absorption proceeds at a faster rate, however the fraction absorbed dose not differ greatly from the value of 0.05 recommended. The excretion in the urine is on the contrary greatly reduced, which could suggest that Ru radionuclides may be retained more efficiently in the body. These differences could have significant impact on the calculation of dose coefficient after incorporation of Ru radionuclides. (author)

  11. Chernobyl fallout in southern and central Finland

    To study the levels and distributions of radionuclides released in the Chernobyl accident, we sampled surface peat from 62 sites in Southern and Central Finland and measured 131I, 134Cs, 137Cs, 132Te, 140Ba, 103Ru, 90Sr, 141Ce, and 95Zr. The distribution of fallout activities was highly uneven, depending on movement of the contaminated air mass and rainfall distribution during the critical days. The highest values observed were 420 kBq m-2 of 131I and 70 kBq m-2 of 137Cs. The nuclide ratios showed wide and partly unexpected variations. The high-boiling-point, or nonvolatile, elements Ce and Zr were spread mostly on a 200-km-wide zone extending across Finland from southwest to northeast. The more volatile elements, I, Ce, and Te, showed quite a different, more widespread, fallout distribution, while an intermediate behavior was observed for Ba, Ru, and possibly Sr. These results can be explained by assuming that pulverized nuclear fuel material released in the reactor explosion on 26 April reached Finland via Poland and the Baltic Sea and traversed the country along the above-mentioned narrow zone, while volatile material, evaporated in the reactor fire from 26 April to 5 May, arrived in several waves and was consequently more widely and evenly spread. From their elemental melting and boiling points, Ru and Mo would appear to belong to the nonvolatile group and Sr to the volatile. Yet, their actual behaviors were opposite; Ru in particular was found in the nonvolatile as well as the volatile fallout, possibly because Ru activities were present in the fuel partly in the metallic state and partly as volatile oxides

  12. Dose assessment for the Metlino and Muslyumovo populations who lived along the Techa river from 1949 to 1954.

    Mokrov, Yuri G

    2004-09-01

    In the period from 1949 to 1956, liquid radioactive waste was routinely and accidentally discharged by the Mayak Production Association, Southern Urals, Russia, into the Techa river. Based on a novel approach, the contamination of the Techa river water, the bottom sediments and the adjacent flood plains was modelled, and internal and external doses were estimated for two villages located downstream of the site of liquid radioactive waste release. Altogether, 11 radionuclides that occurred in the liquid radioactive waste were included in the calculations. The results suggest significantly higher doses than previously assumed, with the major contribution in the year 1951. Radionuclides with half-lives of less than 1 year, such as 89Sr, 131I, 95Zr+95Nb, 103Ru+103mRh, 140Ba+140La, and 144Ce+144Pr, represent the major sources and, in contrast, long-lived radionuclides, such as 90Sr and 137Cs that have so far been assumed to be most important, did not dominate the doses. For adults from the village Metlino, located 7 km downstream of the site of liquid radioactive waste discharge, the committed effective doses due to intake of radionuclides were calculated to be about 2.3 Sv, while the external doses were between about 1.2 Sv and 6.9 Sv. On the other hand, for adults from Muslyumovo, located 75 km downstream, the committed effective doses due to intake of radionuclides were calculated to be about 0.5 Sv, while the external doses were between 0.5 Sv and 2.0 Sv. The values for the skin doses to the Metlino and Muslyumovo populations were about 7.1 Sv and 1.3 Sv, respectively. It is concluded that the current dose estimates for the residents of the Techa river need, therefore, reevaluation. PMID:15378312

  13. Forage: A sensitive indicator for airborne radioactivity

    As a part of the radiological environmental monitoring program at the Joseph M. Parley Nuclear Plant to meet the requirements of NRC Regulations 10 CRF 50, Appendix I, routine sampling of forage was implemented. Indicator plots of forage (grass) were established at the plant site boundary in the two Meteorological sectors having the highest X/Q values for ground-level dispersion of airborne radioactivity. Likewise, a control plot was established in a sector having a significantly lower X/Q value at a distance of 18 miles. Procedures for maintenance of the grass plots, sampling of forage, and sample preparation for measurement of gamma radioactivity with a Ge (Li) detector were developed during the reported three year measurement period. Three atmospheric nuclear tests by the Peoples Republic of China in 1976 and 1977 has proven forage sampling to be convenient, sensitive, and in the judgement of the authors gives results which are superior to most other media sampled for airborne radioactivity. Typical measured levels of radioactivity from 150 to greater than 10,000 pCi/kg (dry weight) were obtained for the principal fission products in the Chinese bomb fallout, which included 95Zr-95Nb, 103Ru, 131I, 140Ba-140La, 141Ce, and 144Ce. On a unit weight basis the level of radioactivity measured was consistently higher for forage than for green leafy vegetables. This was attributed to the higher surface area for the forage. For comparison, plots of airborne concentrations for gross beta and particulate gamma emitters are shown during the time periods that include the Chinese nuclear tests. (author)

  14. Radioactive contamination of bottom sediments in the upper reaches of the Techa river: analysis of the data obtained in 1950 and 1951.

    Mokrov, Yury G

    2003-10-01

    A stationary sorption model has been developed for re-evaluating and analysing archive data from 1950-1951 on the radioactive contamination of Techa river bottom sediments close to the site of liquid radioactive waste discharge. In general, good agreement was obtained between calculations and measurements, which substantiates further the assumptions and conclusions in two preceding articles, on the radionuclide composition of discharged liquid radioactive waste. Estimates on the effective liquid radioactive waste discharges given here are significantly different from those deduced in the 1950s, i.e. in summer 1950 and October 1951. The results are discussed in relation to the Techa River Dosimetry System 2000 (TRDS-2000) that has recently been presented to serve as a means for estimating doses to the Techa river residents. Parameter values describing the exponential decrease of bottom sediment contamination along the river due to short-lived radionuclides, such as (106)Ru, and (144)Ce, agree reasonably with those used in TRDS-2000. However, for other radionuclides, such as (95)Zr, (95)Nb, (91)Y, (90)Sr and (137)Cs, substantial differences are found. It is demonstrated that water flow rate, width of the river, and surface area of bottom sediments are important parameters which were not adequately taken into account in TRDS-2000. Also, the stirring-up of contaminated bottom sediments and their subsequent transport by the water flow are seen to be an important mechanism that governs the radionuclide transport downstream. This mechanism was not included in the TRDS-2000 model. It is concluded that the sorption model used in TRDS-2000 for the reconstruction of radioactive contamination of water and bottom sediments of the Techa river in 1949-1951, is subject to considerable errors. While the present paper is focussed on details of the dosimetric modelling, the implications for the Techa river dosimetry are major. They will be further elucidated in a forthcoming paper

  15. The situation of marine sediments in the pathway of radionuclides to man

    It is widely accepted that the marine sediments are the major reservoir of the radionuclides which are discharged into the coastal sea water. The radionuclides sorpted on the sediments would be recycled in the sea water and be accumulated by marine organisms. And some of the sorpted radionuclides would be transfered to the worm (a kind of benthos) habited in the sediments through their surface or digestive tract and then to the fishes which prey upon the benthos. Therefore, it is important to learn whether the sediments as a ''sink'' may not later become a ''source'' of these contaminants. In this paper, the problem is discussed refering to few reports which are published up to this time. Renfro estimated that about one per cent of 65Zn in the sediment would be transfered to Nereis diversicolor which directly contacts with the sediments: On the recycling of 65Zn from the sediments to the sea water, Seymour and Nelson determined the 65Zn level of 8/100 in the sea water after twelve months following shutdown of reactor. Thus they calculated the ecological half-lives which were affected by the radionuclides retention of the sediments. In our laboratory experiment on the interaction of sediments and Nereis diversicolor using 95Zr, 106Ru, 137Cs and 60Co, the similar results were observed except 137Cs which showed rather high rate of remobilization from the sediments to the sea water. Thus, it could be generally considered that the transfer of radionuclides via sediment would be minor to man. However, it should be born in mind that the increased amount of radionuclides by continuous discharge into sea may raise the level of radionuclides in the sediments and then benthos even with one per cent transfer rate from sediment to benthos. (auth.)

  16. Oceanographic Survey. 3. Radiological survey of marine environment in Ibaraki Prefecture: marine sediment survey

    Radioactivity in marine environment of Ibaraki Prefecture has been monitored since 1965. This report briefly describes the radiological survey data of marine sediments. The concentrations of radionuclides in marine water were determined yearly 4 times for 3H and twice for 54Mn, 60Co, 90Sr, 95Zr, 95Nb, 106Ru, 137Cs and 144Ce. The concentrations of 90Sr, 137Cs and 239+240Pu were hardly changed and those of 90Sr and 137Cs were near the limit of identification. The annual excretion of radionuclides from major nuclear plants in Ibaraki Prefecture including JAERI (Tokai and Oarai), Japan Atomic Power Co., etc. were determined for 3H, 14C, 90Sr, 137Cs, 60Co, 129I and 239+240Pu. Based on the results, dose equivalent for the intake of marine animals and algae was estimated to be as low as 0-4.4 μSv. This level was extremely lower than the limit of annual dose equivalent; 1000 μSv and it was kept nearly constant if the operators of those facilities were not discontinued for a long. To elucidate the factors related to the transfer of radionuclides into marine sediments, the depth dose distribution in marine sediments were investigated as well as the relation between radionuclide level and granular size. An elevation of 137Cs level was found for the sediments composed of fine granules, which were rich in the sediments of offshore region. The 239+240Pu level was positively correlated to the content of organic compounds in the sea. (M.N.)

  17. Implications of radiochemical purity of 99Mo/99Tc generator eluates for the determination of low levels of 99Tc in seawater

    Full text: The determination of sub-Becquerel levels of the long-lived fission product 99Tc in environmental matrices in general and seawater in particular presents analytical challenges, not least with respect to the selection of an appropriate and practicable tracer for calculation of radiochemical yield. Although a number of isotopes (97Tc, 95mTc and 97mTc) have been proposed for this purpose, 99mTc, eluted from easily available 99Mo/99mTc generators, is currently a commonly used tracer due to its availability, convenient assay and practicability. For the analysis of low levels (3 or kg) of 99Tc in seawater samples, attention must be focused on the radiochemical purity of the tracer solution in relation to isotopic contamination with both 99Tc and other radionuclides. Isotopic contamination of eluates from 99Mo/99mTc generators can arise during manufacture and reported impurities include 99Mo, 131I, 132I, 106Ru, 90Sr, 90Y, 89Sr and 103Ru. Of more consequence for the analysis considered here is the presence of 99Tc in such eluates. A cursory examination of the decay scheme of 99Mo indicates that there are two different routes by which 99Tc can be produced within a 99Mo/99mTc generator. Any 99Tc within the eluate will inevitably pass through the analytical sequence and contribute to the final analytical signal. Initial consideration of the problem indicates that correction for the 99Tc contribution is possible knowing the activity and history of the particular generator although the findings of indicate that such procedures may be invalid. To investigate the possible impact of 99Tc contamination on the analysis of low activity seawater samples, a series of investigations were conducted. The generators used in the study were of nominal activity of 25 GBq 99Mo at the time of original calibration and were 2-3 weeks old before use, at which point the 99Mo activity was of the order of 10-20 MBq. Before that time, the generators had been used for their intended radio

  18. Ruthenium release modelling in air under severe accident conditions using the MAAP4 code

    Beuzet, E.; Lamy, J.S. [EDF R and D, 1 avenue du General de Gaulle, F-92140 Clamart (France); Perron, H. [EDF R and D, Avenue des Renardieres, Ecuelles, F-77818 Moret sur Loing (France); Simoni, E. [Institut de Physique Nucleaire, Universite de Paris Sud XI, F-91406 Orsay (France)

    2010-07-01

    In a nuclear power plant (NPP), in some situations of low probability of severe accidents, an air ingress into the vessel occurs. Air is a highly oxidizing atmosphere that can lead to an enhanced core degradation affecting the release of Fission Products (FPs) to the environment (source term). Indeed, Zircaloy-4 cladding oxidation by air yields 85% more heat than by steam. Besides, UO{sub 2} can be oxidised to UO{sub 2+x} and mixed with Zr, which may lead to a decrease of the fuel melting temperature. Finally, air atmosphere can enhance the FPs release, noticeably that of ruthenium. Ruthenium is of particular interest for two main reasons: first, its high radiotoxicity due to its short and long half-life isotopes ({sup 103}Ru and {sup 106}Ru respectively) and second, its ability to form highly volatile compounds such as ruthenium gaseous tetra-oxide (RuO{sub 4}). Considering that the oxygen affinity decreases between cladding, fuel and ruthenium inclusions, it is of great need to understand the phenomena governing fuel oxidation by air and ruthenium release as prerequisites for the source term issues. A review of existing data on ruthenium release, controlled by fuel oxidation, leads us to implement a new model in the EDF version of MAAP4 severe accident code (Modular Accident Analysis Program). This model takes into account the fuel stoichiometric deviation and the oxygen partial pressure evolution inside the fuel to simulate its oxidation by air. Ruthenium is then oxidised. Its oxides are released by volatilisation above the fuel. All the different ruthenium oxides formed and released are taken into consideration in the model, in terms of their particular reaction constants. In this way, partial pressures of ruthenium oxides are given in the atmosphere so that it is possible to know the fraction of ruthenium released in the atmosphere. This new model has been assessed against an analytical test of FPs release in air atmosphere performed at CEA (VERCORS RT8). The

  19. Ruthenium release at high temperature from irradiated PWR fuels in various oxidising conditions. Main findings from the VERCORS program

    Fission product release and transport in case of PWR severe accident is a major topic in reactor safety assessment due to the potential radiological consequences for surrounding populations and the environment. In this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the VERCORS analytical test program which was performed by the ''Commissariat a l'Energie Atomique'' (CEA). It is usually considered as complementary to the PHEBUS FP in-pile integral experimental program. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions).The influence of the nature of the fuel (UO2 versus MOX, burn-up) and the fuel morphology (initially intact or fragmented fuels) have also been investigated. These led to an extended data base allowing on the one hand to study mechanisms which promote fission products release, and on the other hand to enhance models implemented in severe accident codes. Among all the fission products investigated, ruthenium is of specific concern because of its high radiological effects due essentially to the combination of both its short and long half-life isotopes (i.e. 103Ru and 106Ru respectively), but also by its ability to generate volatile gaseous oxides (RuO3, RuO4) in very oxidising conditions, in particular in the case of air ingress accidents. Important uncertainties still remain on the release and transport of this element in such situations, and investigations on this open issue are notably carried out in the SARNET European framework. The present communication gives a general overview of the VERCORS program and presents more deeply the main findings concerning the ruthenium release. Its global behaviour is analysed on the basis of several comparative tests: same UO2 sample (35 and 50 GWd/t) under hydrogen or steam conditions, similar MOX sample (40 GWd/t) under hydrogen

  20. 106Ruthenium Brachytherapy for Retinoblastoma

    Purpose: To evaluate the efficacy of 106Ru plaque brachytherapy for the treatment of retinoblastoma. Methods and Materials: We reviewed a retrospective, noncomparative case series of 39 children with retinoblastoma treated with 106Ru plaques at the Jules-Gonin Eye Hospital between October 1992 and July 2006, with 12 months of follow-up. Results: A total of 63 tumors were treated with 106Ru brachytherapy in 41 eyes. The median patient age was 27 months. 106Ru brachytherapy was the first-line treatment for 3 tumors (4.8%), second-line treatment for 13 (20.6%), and salvage treatment for 47 tumors (74.6%) resistant to other treatment modalities. Overall tumor control was achieved in 73% at 1 year. Tumor recurrence at 12 months was observed in 2 (12.5%) of 16 tumors for which 106Ru brachytherapy was used as the first- or second-line treatment and in 15 (31.9%) of 47 tumors for which 106Ru brachytherapy was used as salvage treatment. Eye retention was achieved in 76% of cases (31 of 41 eyes). Univariate and multivariate analyses revealed no statistically significant risk factors for tumor recurrence. Radiation complications included retinal detachment in 7 (17.1%), proliferative retinopathy in 1 (2.4%), and subcapsular cataract in 4 (9.7%) of 41 eyes. Conclusion: 106Ru brachytherapy is an effective treatment for retinoblastoma, with few secondary complications. Local vitreous seeding can be successfully treated with 106Ru brachytherapy

  1. Environmental radiation measurement in CTBT verification system

    This paper introduces the technical requirements of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) Radionuclide Stations, the CTBT-related activities carried out by the Japan Atomic Energy Research Institute (JAERI), and the ripple effects of such acquired radionuclide data on general researches. The International Monitoring System (IMS), which is one of the CTBT verification regime. Consists of 80 radionuclide air monitoring stations (of those, 40 stations monitor noble gas as well) and 16 certified laboratories that support these stations throughout the world. For radionuclide air monitoring under the CTBT, the stations collect particulates in the atmosphere on a filter and determine by gamma-ray spectrometry the presence or absence of any radionuclides (e.g. 140Ba, 131I, 99Mo, 132Te, 103Ru, 141Ce, 147Nd, 95Zr, etc.) that offer clear evidence of possible nuclear explosion. Minimum technical requirements are stringently set for the radionuclide air monitoring stations: 500 m3/h air flow rate, 24-hour acquisition time, 10 to 30 Bq/m3 of detection sensitivity for 140Ba, and less than 7 consecutive days, or total of 15 days, a year of shutdown at the stations. For noble gas monitoring, on the other hand, the stations separate Xe from gas elements in the atmosphere and, after purifying and concentrating it, measure 4 nuclides, 131mXe, 133Xe, 133mXe, and 135Xe, by gamma-ray spectrometry or beta-gamma coincidence method. Minimum technical requirements are also set for the noble gas measurement: 0.4 m3/h air flow rate, a full capacity of 10 m3, and 1 Bq/m3 of detection sensitivity for 133Xe, etc. On the request of the Ministry of Education, Culture, Sports and Technology, the JAERI is currently undertaking the establishment of the CTBT radionuclide monitoring stations at both Takasaki (both particle and noble gas) and Okinawa (particle), the certified laboratory at JAERI Tokai, and the National Data Center (NDC 2) at JAERI Tokai, which handles radionuclide data, as

  2. Solid phase extraction of actinides using polymeric beads impregnated with TODGA

    137Cs is one of the major radionuclides contributing to the activity of the high level waste (HLW) and its separation facilitates the safe disposal of the latter in deep geological repository as vitrified mass. The global inventory of 137Cs was estimated to be around 3.7x1014 kBq in 2010. Due to its long half-life and reasonable gamma energy (661 keV), 137Cs it has potential application as a radiation source in gamma irradiators in the environmental pollution control, food preservation and sterilization of medical accessories. It is, therefore, required to develop efficient separation methods for its recovery from HLW. We have recently reported that solvent system containing calix(4)arene-crown-6 ligands in FS-13 have yielded much improved extraction of radio-cesium as compared to analogous solvent systems containing nitrobenzene as the diluents. Our studies involving four different calix-bis-crown-6 and calix-mono-crown-6 ligands have indicated that calix(4)arene-dibenzo-bis-crown-6 (CBC) ligands were better suited due to more favourable extraction of Cs(I) and radiolytic stability of the ligand. In the present study, we have tried use the developed solvent system for the recovery of radio-cesium (137Cs and 134Cs) from actual high level waste (HLW). The HLW, containing a total activity of 100 Ci/L also contained 3.85 g/L U and 7.09 mg/L Pu which were removed by contacting the waste solution twice with 30% TBP in n-dodecane in a hot-cell facility. The raffinate from the TBP extraction step contained negligible amounts of the actinides and was subsequently contacted four times with 1.0x10-3 M CBC in FS-13 (each time with fresh organic phase) in the hot-cell facility and the extraction results indicated 137Cs to remain in the raffinate out of 15.88 Ci/L suggesting ∼ 97% extraction in 4 stages. Quantitative extraction can be done by increasing the number of stages. It was interesting to note that none of the other radionuclides, such as 144Ce, 106Ru, 95Zr, 95Nb, 90Y

  3. Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

    Highlights: ► We developed a new modelling of fuel oxidation and ruthenium release in the EDF version of the MAAP4 code. ► We validated this model against some VERCORS experiments. ► Ruthenium release prediction quantitatively and qualitatively well reproduced under air and steam atmospheres. - Abstract: In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO2 to UO2+x. Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence of

  4. Separation and recovery of ruthenium from radioactive liquid waste for specific medical applications - wealth from waste

    In recent past, 106Ru has emerged as one of the promising β- emitting radionuclide used in brachytherapy for the treatment of choroidal melanoma and retinoblastoma due to its favorable nuclear decay characteristics. A plaque with low amount of 106Ru activity of the order of 12 - 26 MBq (0.3 - 0.7 mCi ) is suitable for the above treatment and can be used for an adequate duration of 1-2 years due to suitable half-life (T1/2 = 1.02 y). In order to undertake the preparation of 106Ru plaque, an indigenous availability of this radionuclide with acceptable purity was explored from radioactive liquid waste having wide spectrum of fission products in line with wealth from waste strategy. Process methodology has been developed and standardized at Process Control Laboratory of Waste Immobilization Plant (WIP), Trombay for separation of 106Ru from radioactive liquid waste for intended medical application. (author)

  5. Langzeitergebnisse bei Aderhautmelanom nach 106Ruthenium-Brachytherapie

    Krause, Nona

    2015-01-01

    Introduction: 106Ruthenium-brachytherapy (106Ru-brachytherapy) is an established therapy for small and medium-sized uveal melanomas. The aim of this study was to examine the long-time results in regard to recurrence rate, complication rate, ocular preservation, metastasis rate and survival with malignant uveal and ciliary body melanoma, as well as relevant prognosis factors, subsequent to 106Ru-brachytherapy. Methodology: In this retrospective study of all cases with uveal or with ciliary ...

  6. Effects of the purification techniques applied to industrial wastes on the chemical behaviour of ruthenium 106 in the marine environment. The case of La Hague reprocessing plant releases

    Since nuclear fuel reprocessing started at the La Hague plant in 1966, the efficiency of the purification techniques applied to radioactive wastes has been improved by a factor of 600. The chemical behaviour of 106Ru observed in the liquid releases into the sea and at two coastal sites near the outlet, shows time variations related to the evolution of the reprocessing techniques used. The ''wet chemical method'' based on 106Ru coprecipitation with cobalt sulphide has fulfilled the objectives of environmental protection and the requirements of industrial productivity. As a consequence, 106Ru behaves like the weakly reactive complexes formed during the nitric acid dissolution stage of the process. Since 1989, vitrification of the effluents mainly affected by these complexes and the optimization of the wet method have contributed to the reduction of 106Ru liquid releases into the sea. At the same time, the radionuclide chemical non-reactivity has been attenuated. Measurements of 106Ru activity carried out in the seaweed Fucus serratus show that the availability of nitrosylruthenium in seawater is closely related to hydrolysis of the complexes. Hydrolysis products are the main source of the chemical species available for exchange with the seaweed, its concentration factor being estimated at 500 ± 350. (authors). 26 refs., 10 tabs., 6 figs

  7. Quantitative analysis of the fission product distribution in a damaged fuel assembly using gamma-spectrometry and computed tomography for the Phébus FPT3 test

    (locate and identify the materials and estimate their density with the X-ray tomograms, locate the FP distribution inside the bundle with the gamma emission tomograms) and to automate the processing of the gamma spectra acquired. The specificities of these gamma spectra (high count rate, number of gamma rays, number of measurements, etc.) required in particular to analyse key lines only and needed an original counting loss correction. The method was validated over the pre-test examination of the fuel bundle, through a comparison with the classical gamma analysis method used at the laboratory for objects of known geometry. The final results, given with acceptable uncertainties, gave for all FPs identified (mainly 137Cs, 131I, 132Te, 140Ba, 95Zr, 103Ru, etc.) their quantitative activity profile along the bundle, their retained and released fractions in the bundle, and also some information about their relocation inside the bundle. The results are in very good agreement with other Phébus FPT3 measurements and inventory calculations

  8. Sorption of radionuclides from Pb-Bi melt. Report 1

    Results of laboratory investigations of sorption and interfacial distribution of 54Mn, 59Fe, 60Co, 106Ru, 125Sb, 137Cs, 144Ce, 154,155Eu and 235,238U radionuclides in the system Pb-Bi melt - steel surface are analyzed. It is shown that 106Ru and 125Sb are concentrated in Pb-Bi melt and other radionuclides with higher oxygen affinity are sorbed on oxide deposits on structural materials. Temperature dependences of sorption efficiency of radionuclides are studied. It is shown that there is sharp increase of this value for all radionuclides near the temperature range 350-400 deg C. Recommendations are given on the use of 106Ru and 125Sb as a reference for fuel element rupture detection system with radiometric monitoring of coolant melt samples and 137Cs, 134Cs, 134mCs with radiometric monitoring of sorbing samples

  9. In situ responses of a biological indicator Mytilus edulis L. to the variations of radionuclide discharges from a fuel reprocessing plant

    The introduction of 10 lots of 100kg of mussels in La Hague waters made it possible to investigate the in situ uptake kinetics of 106Ru + 106Rh released by the fuel reprocessing plant of La Hague during 1 year. Equilibrium was reached within 1 - 3 months, whatever the time of introduction and the amounts released. The removal of the mussels labelled at La Hague to a remote aera made it possible to calculate the biological half-life of 106Ru + 106Rh, viz from 16 to 18 days

  10. Behaviour of solute and particle markers in the stomach of sheep given a concentrate diet

    Fistulated sheep given a concentrate diet were used to study the behaviour of solute ([51Cr]EDTA) and particle ([103Ru]phenanthroline) markers in the stomach under conditions of continuous feeding. An injection of a mixed dose of [51Cr]EDTA and [103Ru]phenanthroline was given into the rumen and the time course of marker concentrations in the rumen and the abomasum was recorded. The curves were analysed on the assumption that the stomach of the sheep could be represented as two mixing compartments (reticulo-rumen and abomasum) and a time delay (omasum). This model provided a very good description of the data. [103Ru]-phenanthroline associated with small particles was retained in the rumen much longer than [51Cr]EDTA. Although exchange of [103Ru] phenanthroline occurred between large and small particle fractions, the results suggested that small particles may have been retained somewhat longer in the rumen than solutes. However, it was clear from the results that the mean retention times for particulate matter in the rumen could not be simply obtained using adsorbable markers. Cyclical fluctuations in the concentration of [51Cr]EDTA in the rumen indicated that there were daily variations in net water flux in the rumen. The presence of protozoa was associated with much shorter retention times of both solutes and particles in the rumen. Protozoa were also associated with reduced rumen volumes. (author)

  11. Post Chernobyl-1

    After a description of the sampling and measurement methods, the values of air concentration for 103Ru, 131I, 132Te, 134Cs, 137Cs and 140Ba measured in Sluggia (Italy) from April 29 to May 31, 1986 are here reported. In two filtres 89Sr, 90Sr, 234U, 238Pu and 239-240Pu were also measured

  12. Three-dimensional plume dynamics in the vadose zone: PORFLO-3 modeling of a defense waste leak at Hanford

    In 1973, approximately 450 m3 of liquid containing radioactive and chemical wastes leaked from the 241-T-106 single-shell tank into the vadose zone at the US Department of Energy's Hanford Site in south-central Washington State. The extent of the 137Cs, 144Ce, and 106Ru contaminant plumes in the vadose zone was estimated in 1973 and 1978 by gamma spectrometry in monitoring wells. Using site data and the PORFLO-3 computer model, a three-dimensional, transient plume migration model was developed for 106Ru and 137Cs. The model was calibrated to the 1973 measured plumes of 106Ru and 137Cs. The calibrated model was then used to study plume migration up to 1990. The simulated 106Ru distribution for 1978 extended deeper than reported values. The simulated distribution of 137Cs for 1978 approximated the measured distribution; the 1973 and 1978 137Cs distributions are similar because of the long half-life and high sorption coefficient of 137Cs. 8 figs., 15 refs

  13. Impacts in groundwater of effluents arising in the nuclear industry

    The following topics are discussed: modes of entry of radioactive wastes into groundwater; movement of radionuclides in the ground; degradation processes; behavior of tritium, 90Sr, 137Cs, 106Ru and transuranic elements; potential pathways to man; and impact of releases of radioactive materials to the ground compared to radiation protection standards. (U.S.)

  14. Calculation of beta-ray dose distributions from ophthalmic applicators and comparison with measurements in a model eye

    Dose distributions throughout the eye, from three types of beta-ray ophthalmic applicators, were calculated using the EGS4, ACCEPT 3.0, and other Monte Carlo codes. The applicators were those for which doses were measured in a recent international intercomparison [Med. Phys. 28, 1373 (2001)], planar applicators of 106Ru-106Rh and 90Sr-90Y and a concave 106Ru-106Rh applicator. The main purpose was to compare the results of the various codes with average experimental values. For the planar applicators, calculated and measured doses on the source axis agreed within the experimental errors (106Ru-106Rh and 5 mm for 90Sr-90Y. At greater distances the measured values are larger than those calculated. For the concave 106Ru-106Rh applicator, there was poor agreement among available calculations and only those calculated by ACCEPT 3.0 agreed with measured values. In the past, attempts have been made to derive such dose distributions simply, by integrating the appropriate point-source dose function over the source. Here, we investigated the accuracy of this procedure for encapsulated sources, by comparing such results with values calculated by Monte Carlo. An attempt was made to allow for the effects of the silver source window but no corrections were made for scattering from the source backing. In these circumstances, at 6 mm depth, the difference in the results of the two calculations was 14%-18% for a planar 106Ru-106Rh applicator and up to 30% for the concave applicator. It becomes worse at greater depths. These errors are probably caused mainly by differences between the spectrum of beta particles transmitted by the silver window and those transmitted by a thickness of water having the same attenuation properties

  15. Rapid radiochemical separation of zirconium-95 and niobium-95

    A rapid method for the quantitative separation of 95Zr and 95Nb has been developed. The method is based on the ion flotation of cationic zirconium complex ions with sodium lauryl sulfate (NaLS) from niobium which is masked with hydrogen peroxide. The separation was applied to mixtures of 95Zr and 95Nb initially in oxalic acid solution and quantitative recoveries of the radiochemically pure radioisotopes were obtained. (orig.)

  16. Metabolism and internal decontamination of a radioactive zirconium-niobium salt

    A study in the rat of the metabolism of a salt of 95Zr-95Nb injected by the intravenous route has allowed us to display the following points: - the fast decrease of the blood concentration of radioisotopes, - the durable accumulation of radioisotopes in the bone, - the strong urinary elimination during the first hours. A study of the possibilities of internal decontamination of 95Zr-95Nb was also carried out. Among all the tested chemicals, only zirconium citrate seems effective. (authors)

  17. Study on the environmental behavior of Chernobyl-derived radionuclides in Kyushu Island, Japan

    The environmental behavior of Chernobyl-derived radionuclides in Kyushu Island was investigated for one month after the accident. The radioactivity level in airborne dusts was two orders of magnitude lower than that observed in Western Europe. The distribution of 131I in airborne dusts shifted to a larger particle size compared with other radionuclides. The radionuclide concentration in seaweeds varied depending on the geographical situation where the sampling was done. The biological half-lives in red algae were calculated to be 17.4 d and 32.9 d for 131I and 103Ru, respectively. The concentration factors in red algae were estimated to be 3 x 103 and 5 x 103 for 131I and 103Ru, respectively. The cooking effect of 131I in seaweeds and the committed effective dose equivalent through ingestion of seaweed were also evaluated. (author)

  18. Pyrolysis of TBP waste with synthetic mica

    One method for treatmenting radioactive waste solvents from a spent fuel reprocessing plant is to convert them to solid inorganic products for stable long-term storage. This study examines the pyrolysis of waste tri- butyl phosphate (TBP) with synthetic mica compound using radioactive tracers and measuring the radioactive tracers retention in the stratiform structure of the synthetic mica pyrolysis product. Cold testing was performed with pure TBP, and hot testing was performed with 103Ru, 131I, 125Sb and 137Cs tracers. The pyrolysis product was composed of stable compounds with nearly complete adsorption of 103Ru, 125Sb and 137Cs tracers. The decomposed TBP waste was present as a phosphate

  19. ZZ MATXSLIBJ33, JENDL-3.3 based, 175 N-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group

    1 - Description of program or function: JENDL-3.3 based, 175 neutron-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group transport codes. Format: MATXS. Number of groups: 175 neutron, 42 gamma-ray. Nuclides: 337 nuclides contained in JENDL-3.3: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Ge-70, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-79, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-87, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-91, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-106, Rh-103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-110m, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127m, Te-128, Te-129m, Te-130, I-127, I-129, I-131, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-130, Ba-132, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-138, La-139, Ce-140, Ce-141, Ce-142, Ce-144, Pr-141, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-148m, Pm-149, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Eu

  20. ZZ MCB-JEF2.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K

    1 - Description of program or function: MCB-JEF2.2 is a continuous-energy cross section libraries in ACE Format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different Temperatures, and cover the energy range up to 20 MeV. Format: ACE. Number of groups: Continuous energy. Nuclides: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat., N-14, N-15, O-16, O-17, Na-23, F-19, Mg-nat., Al-27, Si-nat., P-31, S-32, S-33, S-34, S-36, Cl-nat, K-nat, Ca-nat., Ti-nat, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-59, Ni-60, Ni-61, Ni-62, Ni-64, Cu-nat, Ga-nat, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-86, Rb-87, Sr-84, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-90, Y-91, Zr-nat, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-nat, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-105, Ru-106, Rh-103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-111, Cd-nat., Cd-106, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115, Cd-116, In-113, In-115, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-24, Sn-125, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Sb-126, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127, Te-128, Te-129, Te-130, Te-132, I-127, I-129, I-130, I-131, I-135, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-139, La-140, Ce-140, Ce-141, Ce-142, Ce-143, Ce-144, Pr-141, Pr-142, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-149, Pm-151, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu

  1. Radioactivity size distributions of ambient aerosols in Helsinki, Finland during May 1986 after Chernobyl accident

    Ambient aerosol size distributions oof 131I, 103Ru, 132Te and 137Cs radionuclides were measured in Helsinki, Finland during May 7 - 14, 1986. Radioactivity size distributions were unimodal. Geometric mean diameter of 131I was in the size range 0.33 - 0.57 μm a.e.d.. Other isotopes had geometric mean diameters in the size range 0.65 - 0.93 μm a.e.d.. (author)

  2. Fission product release from ZrC-coated fuel particles during postirradiation heating at 1600 C

    Release behavior of fission products from ZrC-coated UO2 particles was studied by a postirradiation heating test at 1600 C (1873 K) for 4500 h and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the test. Ceramographic observations showed no palladium attack and thermal degradation of ZrC. Fission products of 137Cs, 134Cs, 106Ru, 144Ce, 154Eu and 155Eu were released from the coated particles through the coating layers during the postirradiation heating. Diffusion coefficients of 137Cs and 106Ru in the ZrC coating layer were evaluated from the release curves based on a diffusion model. 137Cs retentiveness of the ZrC coating layer was much better than that of the SiC coating layer. ((orig.))

  3. 4. Quarterly progress report, 1982

    This quarterly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 58Co, 60Co, 134Cs, 137Cs, 125Sb, 90Sr, 106Ru, K, 54Mn, U and T). A bibliographic selection is also presented

  4. Separation of valuable fission products from High Level Liquid Waste (HLLW)

    High Level Liquid Waste (HLLW) generated during spent fuel reprocessing (PUREX) is an important source of valuable fission products viz. 137Cs, 90Sr (and its daughter product 90Y), platinum group metals like 106Ru , Pd etc. We present a short overview of the work carried out in our laboratory on the separation of the valuable radionuclides from actual research reactor HLLW employing a host of separation techniques like solvent extraction, precipitation, liquid membrane etc. (author)

  5. Partitions of 4d Transition Metal Nuclei and Related Correlations Using the Core Cluster Model

    Mageed, K E Abd El

    2013-01-01

    In the present work we attempt to study the cluster model in the transition metal region. The spectrum fitting method is studied for the selected nuclei (88,90,92^Sr, 92,94^Zr, 98,100^Mo, 100,102,104, 106^Ru, 108,110^Pd and 112,114,116,118^Cd) with proton number (38 0+), the excitation energies and the product of valence nucleon numbers of the parent nuclei.

  6. 1. Quaterly progress report, 1980

    This quaterly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are in tables (7Be, 131I, 137Cs, 90Sr, 106Ru and 106Rh, 226Ra, 54Mn, U and T). A bibliographic selection is also presented

  7. Preparation of ruthenium nitrosyl nitrate solutions and the extraction of ruthenium complexes by means of neutral organic phosphorus compounds

    A method is described for the preparation of solutions of nitrosyl nitrate complexes of ruthenium. Ruthenium chloride has been transformed into ruthenium tetroxide which reacted with nitrogen oxides in nitric acid and formed ruthenium nitrosyl nitrate. The solution has been labelled with 106Ru and spectrophotometrically investigated. The extraction behavior of the ruthenium complexes with neutral organic phosphorous compounds has been investigated as a function of the acid and salt concentration of the solution and compared with that of other fission products. (author)

  8. Application of multiple gamma-ray spectrum for analytical chemistry

    Hatsukawa, Yuichi; Hayakawa, Takehito; Shinohara, Noboru; Oshima, Masumi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    Feasibility of application of the multi-gamma ray spectrum for analytical chemistry was examined. A specimen in which some minor fission products are included was measured at an array of ten germanium detectors with BGO Compton suppressors, GEMINI, and multiple gamma-ray spectra are measured. Even in very strong radiation fields from {sup 137}Cs isotope, some miner contents, {sup 106}Ru, {sup 125}Sb, {sup 144}Pr, {sup 207}Bi were detected by this method. (author)

  9. 2. Quarterly progress report, 1983

    This quarterly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 58Co, 60Co, 134Cs, 137Cs, 90Sr, 106Ru, K, 54Mn, U and T). A bibliographic selection is also presented

  10. 3. Quarterly progress report 1982

    This quarterly report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, irrigation water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 58Co, 60Co 134Cs, 137Cs, 90Sr, 106Ru, K, 54Mn, U and T). A bibliographic selection is also presented

  11. Acute toxicity of beta-emitting radionuclides that may be released in a reactor accident and ingested

    Suckling, weanling, and adult rats received 106Ru--106Rh by gavage and adult beagle dogs ingested 106Ru--106Rh to determine the toxicity of this high-energy (1.4 MeV av) β-emitting nuclide pair. The LD50's for suckling, weanling, and adult rats were 1.5, 18, and 9.0 mCi/kg, respectively. Adult rats were given 147Pm by gavage to determine if a low-energy (0.06 MeV av) β emitter could also cause death by damaging the bowel. The LD50 of 147Pm in rats was about 5 Ci/kg. The calculated radiation doses absorbed in the target cells at the LD50 level were approximately the same for the two radionuclides (3500 rad), although the doses at the mucosal surface differed widely. The LD50 for 106Ru--106Rh in dogs was about 3.5 mCi/kg. Dosimeters placed beneath the mucosa in dogs indicated that the radiation dose to the target cells that caused death from 106Ru--106Rh was about the same as it was for rats. The signs of intestinal injury, their duration, and the probabilities of tissue repair were much different in the dog than in the rat. The midcolon and lower colon of dogs were usually denuded at focal sites after ingestion of 2.5 to 4.0 mCi/kg, and frequently that damage was irreversible. The fatal consequence of severe mucosal damage was averted in two dogs by colectomy and an ileorectal anastomosis

  12. Placental transfer of ruthenium in rat and guinea-pig

    Ruthenium-106 in citrate solution was administered intravenously to rat at different stages of pregnancy and to guinea-pig either before conception or in late pregnancy. The results for rat showed that retention in the embryo/foetus measured at 3-5 days after administration increased from about 0.0002% of injected activity per embryo/foetus on day 12 of gestation to about 0.05% at birth. The relative concentrations of 106Ru in embryo/foetus and mother (Cf/Cm ratio) were about 0.1 in each case. Concentrations in the yolk sac on day 12 were about 1% g-1 compared with 0.01% g-1 kin the foetus/ Retention in the guinea-pig foetus in late gestation at 7 days after administration (days 50-57) was about 0.2% injected activity per foetus, corresponding to a Cf/Cm = 0.2. Retention in each foetoplacental unit was 2% of injected 106Ru with 50% in the yolk sac, 35% in the placenta and 10% in the foetus. For administration 4 weeks prior to conception, the level of 106Ru retained in the foetus on day 57 of gestation was two orders of magnitude lower than after short-term administration, with a Cf/Cm about 0.004. (author)

  13. Combining of some trace elements with constituent materials of marine algae

    Two radionuclides (137Cs and 106Ru-106Rh) were extracted from a brown alg a (Eisenta bicyclis) into 5 solvents (Ethyl ethel, 80% Ethyl alcohol, boiled water, 0.2% NaOH and 24% KOH) in different proportions, suggesting that both radionuclides do not combine with fats and pigments, and that 137Cs associates maybe with dextrans and monosaccharides, while, 106Ru-106Rh mainly combines with the cell wall polysaccharides such as alginic acid and fucoidan. In order to obtain information from extracts of algae, gel filtration was carried out on 2 species of algae (Ulva pertusa and Eisenia bicyclis) using Sephadex G-100 and G-25. Gel filtration profile gave only one peak for 137Cs, 2 for 106Ru-106Rh and 125I, and 3 for 60Co corresponding to positions where saccharides of the algae appeared. As the result, it was found that different radionuclides combined with different constituent materials of an alga, to some extent. Gel filtration profiles of 125I were compared with each other among several species of marine algae. They were different from one another among classes of green, brown and red algae, though they were similar in a class. Gel filtration profiles of 125I were also varied between 2 chemical forms of 125I (Na125I and Na125IO3). (J.P.N.)

  14. Development of A phantom for ophthalmic beta source applicator quality control using TL dosimetry

    Barbosa, N. A.; da Rosa, L. A. R.; Braz, D.

    2015-11-01

    Concave eye applicators with 90Sr/90Y and 106Ru/106Rh beta ray sources are usually used in brachytherapy for the treatment of superficial intraocular tumors as uveal melanoma with thickness up to 5 mm. The calculation of the dose delivered to the eye is carried out based on the data present in the beta source calibration certificate. Therefore, it would be interesting to have a system that could evaluate that dose. In this work, an eye phantom to be used with 106Ru/106Rh betatherapy applicators was developed in solid water. This phantom can hold nine micro-cube thermoluminescent (TL) dosimeters, TLD-100. The characteristics of the TL response of the dosimeters, namely reproducibility and individual sensitivity, were determined for a 60Co source. Using Monte Carlo code MCNPX, the dose to a water eye was determined at different depths. Exposing the eye phantom with TL dosimeters to the 106Ru/106Rh applicator, it is possible to assess calibration factors using the dose values obtained by Monte Carlo simulation to each depth. Using mean calibration factors, dose values obtained by TL dosimetry were compared to the data present in the applicators certificate. Mean differences for both applicators were lower than ±10%, maximum value 17% and minimum value 0.08%. Considering that the certificate values present an uncertainty of ±20%, the calibration procedure and the developed phantom are validated and can be applied.

  15. Brachytherapy treatment simulation of strontium-90 and ruthenium-106 plaques on small size posterior uveal melanoma using MCNPX code

    Abstracts: Concave eye applicators with 90Sr/90Y and 106Ru/106Rh beta-ray sources are usually used in brachytherapy for the treatment of superficial intraocular tumors as uveal melanoma with thickness up to 5 mm. The aim of this work consisted in using the Monte Carlo code MCNPX to calculate the 3D dose distribution on a mathematical model of the human eye, considering 90Sr/90Y and 160Ru/160Rh beta-ray eye applicators, in order to treat a posterior uveal melanoma with a thickness 3.8 mm from the choroid surface. Mathematical models were developed for the two ophthalmic applicators, CGD produced by BEBIG Company and SIA.6 produced by the Amersham Company, with activities 1 mCi and 4.23 mCi respectively. They have a concave form. These applicators' mathematical models were attached to the eye model and the dose distributions were calculated using the MCNPX ⁎F8 tally. The average doses rates were determined in all regions of the eye model. The *F8 tally results showed that the deposited energy due to the applicator with the radionuclide 106Ru/106Rh is higher in all eye regions, including tumor. However the average dose rate in the tumor region is higher for the applicator with 90Sr/90Y, due to its high activity. Due to the dosimetric characteristics of these applicators, the PDD value for 3 mm water is 73% for the 106Ru/106Rh applicator and 60% for 90Sr/90Y applicator. For a better choice of the applicator type and radionuclide it is important to know the thickness of the tumor and its location. - Highlights: ► 106Ru and 90Sr β applicators were modeled using Monte Carlo code MCNPX. ► Dose distributions were calculated for all eye structures, including a tumor region. ► 106Ru generates higher lens doses than those generated by 90Sr

  16. ZZ FSXJ32, MCNP nuclear data library based on JENDL-3.2. ZZ FSXLIBJ33, MCNP nuclear data library based on JENDL-3.3

    1 - Description of program or function: - NEA-1424/03: JENDL-3.2 based MCNP library. Format: MCNP. Number of groups: Continuous energy cross section library. Nuclides: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Hf, Ta, W, Pb, Bi, Ra, Ac, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es, Fm. Temperatures: 293 K, 600 K, 900 K, 1200 K, 1500 K, 2000 K. Origin: JENDL-3.2. The temperature-dependent continuous energy cross section library for MCNP, FSXJ32, was prepared from JENDL-3.2 for a variety of applications in the field of atomic energy. - NEA-1424/06: April 2005: This is the DVD version of ZZ-FSXJ32 NEA-1424/03. - NEA-1424/07: This version differs from version NEA-1424/05 in the following: Index files xsdir.fsxlb331 and xsdir.fsxlb332 have been updated, since atomic weights were missing for 23 nuclides. JENDL-3.3 based MCNP library. Format: MCNP. Number of groups: Continuous energy cross section library. Nuclides: 337 nuclides contained in JENDL-3.3. H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Ge-70, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-79, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-87, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-91, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-106, Rh

  17. High-Performance γ spectrometry Using Ge(Li) Detectors

    This report describes a high resolution gamma spectrometer design which use Ge-Li detectors, a cooled field effect transistor preamplifier, and a spectrum stabiliser. The obtained resolution and the 122 keV gamma ray of the 57Co is 0.96 keV, and 239Pu, 233Pa and 95Zr + 95Nb spectra are shown for the example. (authors)

  18. Metabolism and internal decontamination of a radioactive zirconium-niobium salt; Metabolisme et decontamination interne chez le rat, d'un sel de zirconium-niobium radioactif

    Gavend, M.; Rinaldi, J.; Rinaldi, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1969-07-01

    A study in the rat of the metabolism of a salt of {sup 95}Zr-{sup 95}Nb injected by the intravenous route has allowed us to display the following points: - the fast decrease of the blood concentration of radioisotopes, - the durable accumulation of radioisotopes in the bone, - the strong urinary elimination during the first hours. A study of the possibilities of internal decontamination of {sup 95}Zr-{sup 95}Nb was also carried out. Among all the tested chemicals, only zirconium citrate seems effective. (authors) [French] Une etude chez le rat du metabolisme d'un sel de {sup 95}Zr-{sup 95}Nb administre par voie veineuse a permis de mettre en evidence les points suivants: - une diminution rapide de la concentration sanguine en radioelements, - une accumulation durable des radioelements dans les os, - une forte excretion urinaire durant les premieres heures. Une etude des possibilites de decontamination interne du {sup 95}Zr-{sup 95}Nb a egalement ete poursuivie. De tous les composes chimiques qui ont ete testes, seul le citrate de zirconium semble etre efficace. (auteur)

  19. Extraction of nitro complexes of ruthenium nitrosyl by quaternary ammonium salt

    103Ru-labeled Na2[Ru(NO2)4OH] complex is prepared and its extractive properties studied in quaternary ammonium-xylene. Distribution coefficients are as high as 103. There are two molecules of quaternary ammonium in the extraction complex. The behaviour and states of ruthenium in the nitric acid solution of irradiated nuclear fuels are investigated. The method of separation and determination of ruthenium in fission products with quaternary ammonium extraction is established. The specific activity of ruthenium in the nuclear fuels is in agreement with the value measured by distilling ruthenium with H2SO4-NaBiO3. (author)

  20. Liquid-liquid extraction of carrier-free radioisotopes produced in α-particle activated molybdenum target by HDEHP and TBP

    Simultaneous production of carrier free isotopes like 95,96Nb, 93,94,95,96,99mTc and 94,95,97,103Ru through the nuclear reactions (α, αpxn), (α pxn) and (α, xn) has been performed by α-particle activation of a molybdenum target. The sequential separation of the produced radioisotopes from the bulk target matrix has been achieved through LLX using HDEHP and TBP as liquid exchangers. Formation of the carrier free radionuclides in the target matrix and their purity in different stages of separation have been verified by taking recourse to γ ray spectrometry. (Author)

  1. Study of the radfioisotopic composition of rain in 2.05 1986 in Bucharest-Magurele area

    The paper presents the activities of radionuclides in the rain that fell in Bucharest-Magurele area on May 2nd, 1986. The artificial radionuclides measured were: Sr-89, Sr-90, Ru-103, Ru-106, Sb-125, I-131, (I+Te)-132, Cs-134, Cs-137, (Ba+La)-140, Ce-141, Ce-144, originating in the nuclear accident at Chernobyl. The activities of I-131 and Cs-137 were 28700 Bq/m2 and 3610 Bq/m2 respectively. (authors)

  2. Brachytherapy treatment simulation of strontium-90 and ruthenium-106 plaques on small size posterior uveal melanoma using MCNPX code

    Barbosa, N. A.; da Rosa, L. A. R.; Facure, A.; Braz, D.

    2014-02-01

    Concave eye applicators with 90Sr/90Y and 106Ru/106Rh beta-ray sources are usually used in brachytherapy for the treatment of superficial intraocular tumors as uveal melanoma with thickness up to 5 mm. The aim of this work consisted in using the Monte Carlo code MCNPX to calculate the 3D dose distribution on a mathematical model of the human eye, considering 90Sr/90Y and 160Ru/160Rh beta-ray eye applicators, in order to treat a posterior uveal melanoma with a thickness 3.8 mm from the choroid surface. Mathematical models were developed for the two ophthalmic applicators, CGD produced by BEBIG Company and SIA.6 produced by the Amersham Company, with activities 1 mCi and 4.23 mCi respectively. They have a concave form. These applicators' mathematical models were attached to the eye model and the dose distributions were calculated using the MCNPX *F8 tally. The average doses rates were determined in all regions of the eye model. The *F8 tally results showed that the deposited energy due to the applicator with the radionuclide 106Ru/106Rh is higher in all eye regions, including tumor. However the average dose rate in the tumor region is higher for the applicator with 90Sr/90Y, due to its high activity. Due to the dosimetric characteristics of these applicators, the PDD value for 3 mm water is 73% for the 106Ru/106Rh applicator and 60% for 90Sr/90Y applicator. For a better choice of the applicator type and radionuclide it is important to know the thickness of the tumor and its location.

  3. Hodoscope module with miniature photomultipliers

    The experimental Scintillation Magnetic Spectrometer (SMS) installation, whose main element is an extended hodoscope system, is being built for the accelerator of the High Energy Laboratory of the Joint Institute for Nuclear Research. The authors describe the scintillation hodoscope of the SMS installation and present the applicable amplitude and time characteristics of several types of miniature photomultipliers (FEU-58, FEU-60, FEU-114-1, FEU-147-1, and R-1635 (Hamamatsu, Japan)), which were obtained with a 106Ru radioactive source and standard plastic scintillators of two types, based on oxazoles in polystyrene and in polymethylmethacrylate

  4. Recovery of nitric acid from simulated acidic high level radioactive waste using pore-filled anion exchange membranes

    Acidic waste is generated at different stages of nuclear fuel cycle. The waste contains minor amounts of actinides (241Am, Pu, Np) along with large number of long-lived radionuclides such as 137Cs, 90Sr, 106Ru etc. Before disposal or storage, the overall activity of the waste needs to be reduced. Along with this, the high amount of acid present in the waste needs to be removed. In this study, DD has been used to recover nitric acid from acidic solutions with compositions similar to radioactive waste using pore-filled anion exchange membranes

  5. Measurement of leaching from simulated nuclear-waste glass using radiotracers

    The use of radiotracer spiking as a method of measuring the leaching from simulated nuclear-waste glass is shown to give results comparable with other analytical detection methods. The leaching behavior of 85Sr, 106Ru, 133Ba, 137Cs, 141Ce, 152Eu, and other isotopes is measured for several defense waste glasses. These tests show that radiotracer spiking is a sensitive, multielement technique that can provide leaching data, for actual waste elements, that are difficult to obtain by other methods. Additionally, a detailed procedure is described that allows spiked glass to be prepared with a suitable distribution of radionuclides

  6. PMT signal increase using a wavelength shifting paint

    We report a 1.65 times increase of the PMT signal and a simple procedure of application of a new wavelength shifting (WLS) paint for PMTs with non-UV-transparent windows. Samples of four different WLS paints, made from hydrocarbon polymers and organic fluors, were tested on a 5-in. PMT (ET 9390KB) using Cherenkov radiation produced in fused silica disks by 106Ru electrons on a ‘table-top’ setup. The best performing paint was employed on two different types of 5-in. PMTs (ET 9390KB and XP4572B), installed in atmospheric pressure CO2 gas Cherenkov detectors, and tested using GeV electrons

  7. Review of literature on bioassay methods for estimating radionuclides in urine

    Bioassay methods of certain important radionuclides encountered in the nuclear fuel cycle operations, viz., thorium, uranium, sup(239)Pu, sup(241)Am, sup(90)Sr, sup(99)Tc, sup(106)Ru, sup(137)Cs are reviewed, with special emphasis on urinalysis. Since the preconcentration is an important prerequisite for bioassay, various preconcentration methods are also discussed. Brief account of various instruments both nuclear and analytical used in the bioassay programme is included. The sensitivities of the methods cited in the literature vis-a-vis the derived recording levels indicated in ICRP recommendations are compared. Literature surveyed up to 1990 is tabulated. (author). 96 refs., 1 fig ., 3 tabs

  8. New results from an imaging CCD used as a position sensitive detector at standard TV rate and room temperature

    This paper shows new and improved results, with respect to previous work, regarding the use of an area imaging CCD (Fairchild CCD 222 running at standard TV rate and at 296 K) as X-ray and charged particle detector. A more dedicated video processor is used and as a consequence single pixel signals, revealing X-rays from a 55Fe source and β-rays from a 106Ru source, can be processed, on the basis of the dark average response and of the rms noise of each pixel. Tests are reported concerning the correlation of the signals with the dark current pattern and with the low light level illumination. (orig.)

  9. Solidification and storage of Savannah River Plant radioactive waste

    A conceptual process for solidification of SRP aqueous waste into high-integrity, low-leachable forms was developed. The process separates about 99.9 percent of the biological hazard from approximately 97 percent of the waste. The major biological hazard remaining in the residual salt is 106Ru, which has a 369-day half-life. Sludge and zeolite can be solidified into concrete or glass. Cost-risk analyses are being made to determine which of these forms is preferred for SRP waste. Also, studies on other parts of the conceptual process are in progress. (auth)

  10. Behaviour of solid fission products in the HTGR coated fuel particles

    Results of profile measurements for volume concentrations of 134,137Cs, 144Ce, 155Eu, 106Ru and fissionable material in the HTGR coated fuel particles which have been subjected to standard tests in the temperature range of 1273-2133 K and at burnup up to 17% fima are presented. Values of the effective coefficients of cesium diffusion in kern and protective coating of fuel particles which were subjected to standard in-pile tests in spherical fuel elements at the temperature of 1273 K and the burnup up to 15% fima as well as the value of relative release of solid fission products from the samples studied are given

  11. Proposed Search for a Fourth Neutrino with a PBq Antineutrino Source

    Several observed anomalies in neutrino oscillation data can be explained by a hypothetical fourth neutrino separated from the three standard neutrinos by a squared mass difference of a few eV2. We show that this hypothesis can be tested with a PBq (ten kilocurie scale) 144Ce or 106Ru antineutrino beta source deployed at the center of a large low background liquid scintillator detector. In particular, the compact size of such a source could yield an energy-dependent oscillating pattern in event spatial distribution that would unambiguously determine neutrino mass differences and mixing angles.

  12. 1. Quarterly progress report 1984

    This report of the SCPRI exposes an interpretation of the principal results concerning the surveillance of radioactivity in the environment: atmospheric dusts, rainwater, surface water, underground water, sewage water, drinking water, food chain, sea water around nuclear plant sites and other sites. The activities of various radioisotopes are presented in tables (7Be, 54Mn, 58Co, 60Co, 90Sr, 95Nb, 106Ru, 110Ag, 125Sb, 131I, 134Cs, 137Cs, 144Ce, 226Ra, U, K, T and Rn). This report exposes also the state of surveillance and assistance operations on work sites and, the state of incidents along the three months; a bibliographic selection is also presented

  13. PMT signal increase using a wavelength shifting paint

    Allada, K; Ou, L; Schmookler, B; Shahinyan, A; Wojtsekhowski, B

    2015-01-01

    We report a 1.65 times increase of the PMT signal and a simple procedure of application of a new wavelength shifting (WLS) paint for PMTs with non-UV-transparent windows. Samples of four different WLS paints, made from hydrocarbon polymers and organic fluors, were tested on a 5-inch PMT (ET 9390KB) using Cherenkov radiation produced in fused silica disks by $^{106}$Ru electrons on a `table-top' setup. The best performing paint was employed on two different types of 5-inch PMTs (ET 9390KB and XP4572B), installed in atmospheric pressure CO$_2$ gas Cherenkov detectors, and tested using GeV electrons.

  14. Report on the intercomparison run IAEA-308 radionuclides in seaweed mixture

    The results of an intercomparison exercise on a sample of mixed seaweeds from the Mediterranean Sea, IAEA-308, designed for the determination of artificial and natural radionuclide levels, are reported. The data from 67 laboratories representing 33 countries have been evaluated. The recommended median values, with confidence intervals, for the most frequently measured radionuclides 106Ru, 110mAg, 134Cs, 137Cs, 238Pu, 239+240Pu, 241Am, 40K, 210Pb and 228Th are given. Refs and tabs

  15. Report on the intercomparison run IAEA-307 radionuclides in sea plant

    The results of an intercomparison exercise on a Mediterranean Sea plant sample, coded as IAEA-307, designed for the determination of artificial and natural radionuclide levels, are reported. This sample was collected along the shore from the vicinity of the Principality of Monaco in October 1986. The data from 66 laboratories representing 31 countries have been evaluated. The most frequently measured radionuclides: 106Ru, 110mAg, 134Cs, 137Cs, 238Pu, 239+240Pu, 241Am, 40K and 226Ra. 10 refs, 38 tabs

  16. Programs and procedures for assessing quality of spectral gamma-ray borehole data for the UGTA

    This report describes the procedures and computer programs used to process spectral gamma-ray borehole logging data in the UGTA (UnderGround Test Area) program at the NTS (Nevada Test Site) to assess data quality. These programs and procedures were used to analyze data from five boreholes in the UGTA program. Development of these computer programs and procedures required considerable effort and the primary purpose of this report is to provide continuity with future activities related to spectral gamma-ray borehole logging in the UGTA program. This is especially important because of the long time interval between cessation of logging in April, 1996 and the next round of activity, which has not yet occurred. This report should also be useful if any quality control issues arise regarding past or forthcoming spectral gamma-ray log analyses. In the characterization work underway at the Nevada Test Site Underground Test Area, the logging contractor, Western Atlas, agreed to identify five artificial nuclides based on their gamma-ray signatures. Those nuclides are 60Co, 106Ru, 125Sb, 134Cs, and 137Cs. In the case of 106Ru, which is not a gamma emitter, any detected gamma rays come from the daughter nuclide 106Rh which has a half-life of 30 s. With such a short half-life, 106Rh can be considered to be in equilibrium with 106Ru under most conditions so the result is the same as if the gamma rays were emitted by the 106Ru. The Western Atlas spectral gamma-ray curve plots from a given borehole present detailed qualitative information on the apparent distribution of natural and artificial nuclides with depth in the borehole. The computer programs and procedures described in this report were used to provide a quality analysis of the contractor's processed data and to work with the contractor to validate and/or refine their existing automatic processing. This was done using a procedure that was developed and tested successfully in earlier work at the NTS; the revised and updated

  17. Deposition of long-lived radionuclides after the Chernobyl accident in the forestal massif of Boreon

    After the reactor accident at Chernobyl, samples of soil, moss, lichen and fern were collected in the forest around the Vesubie valley in the South East fo France and analyzed by low energy photon and gamma spectrometry. Activity concentrations as high as 42.8, 9.4 and 3.8 kBq.m-2 were measured for 137Cs, 134Cs and 106Ru, respectively, in soil, in October 1988. 12Sb and 110mAg were also detected. The contamination was found to be the most important between 1400 and 1700 m altitude. (author) 9 refs.; 5 figs.; 2 tabs

  18. Silica aerogel Cerenkov detectors for particle identification

    We present light yield measurements of silica aerogel Cerenkov detectors with photomultiplier readout, showing the light yield dependence of pure and wavelength-shifter-doped silica aerogel on block size using both cosmic muons and electrons from a 106Ru source. We present studies of fluorescent fibers and single photon avalanche diodes, including measurements of attenuation lengths and emission spectra of fibers versus wavelength and tests with a single photon avalanche diode. We show results of the response of a single photon avalanche diode to different light sources. Finally, we discuss a new readout scheme using avalanche photodiodes

  19. Dynamics of contents and organic forms of radionuclide compounds in the liquid phase of forest soils in the zone of contamination from the Chernobyl Nuclear Power Plant

    In the profile of forest soils in a 30-km zone around the Chernobyl nuclear power plant (NPP), in areas characterized by different positions in relation to the source of emission, the authors determined the relative contents of long-lived radionuclides 106Ru, 134,137Cs, and 144Ce in soil solutions (as of 1987). On the example of 137Cs, they consider the dynamics (1987-1990) of relative contents and forms in which the radionuclide is found in the liquid phase of soils in the zone of radioactive contamination from the Chernobyl NPP

  20. Silviculture-ecological consequences of forest pollution due to radioactive effluents

    Radioactive contamination effect on the forest areas of Pripyat' Polessie is considered. Radiation processes in damaged pinetree plantations are characterized. Radionuclide migration dependent on soil types and tree stocks is analyzed. The data analysis has shown the evidence of 144Ce, 1'37Cs, 134Cs, 106Ru in 3 years after radioactive contamination in the controlled area. By the end of the third year a significant radionuclide migration had occurred between the forest floor and lower aquifers. refs. 2; figs. 2; tabs. 8

  1. Nuclear moments of transitional nuclei in the vicinity of Z=40 and Z=82 shells

    With the NICOLE facility on-line to the isotope separator ISOLDE-3 at CERN we obtained on the gyromagnetic factors of 182,184m,184gAu using the nuclear spin-lattice relaxation phenomena; this new method uses a radioactive pulsed beam and a time resolved detection. We developed an algorithm based on the relaxation theory which allows to calculate the time evolution of the angular distribution of γ-radiation emitted in radioactive decay of oriented nuclei. The possibilities for the 182Au spin value were reduced to I=2,3,4; its half-life was reevaluated to T1/2(182Au) = 15.6(4)s. The first on-line experiment carried out at NICOLE on 184Au showed the existence of an unknown long-lived isomeric state which decays to the short-lived ground state through an M3 transition. Our results indicate that the most probable spin sequence is 3+ and 6+ for the isomeric and ground state, respectively; the half-lives are T1/2 (184mAu) = 48(1)s and T1/2 (184Au)=19(1)s. Magnetic moments of 91mY, 95Zr and 97Nb and quadrupole moments of 95Zr and 95Nb have been measured using the time-integral nuclear orientation and the nuclear magnetic resonance methods at low temperature. The measured magnetic moment |μ (95Zr)| =1.131(20)μN gives a quadrupole deformation ε 0.06 - 0.07 for 95Zr using the Nilsson model with a Coriolis mixing term. This deformation explains the observed value for the quadrupole moment Q(95Zr) = +0.29(5) eb. The small oblate deformation of 95 Nb is in agreement with the spectroscopic quadrupole moment Q (95Nb) = -0.28(10) eb. (author). 120 refs., 43 figs., 13 tabs

  2. Long-lived radionuclide-impurities in eluates from molybdenum-technetium generators and the associated absorbed dose to the patient

    he activity-concentrations of several long-lived gamma-emitting radionuclides present in technetium generators and in eluates from these generators have been determined by means of Ge(Li) gamma-spectrometry. The principal contaminants of the eluates were: 192Ir(T sub (1/2) = 74.3 d), 134Cs (2.05 a), 131I (8.05 d), 110Ag sup (m) (255 d), 103Ru (39.5 d), 99Mo (66.7 h) and 60Co (5.26 a). Thhe impurity-concentrations were found to vary considerably from generator to generator. Changes in the impurity-concentrations in eluates from the same generator have also been recorded during an elution-period of one week. In accord with their ability to be eluted from the generators, the long-lived radionuclide-impurities may be arranged in the following sequence 134Cs > 103Ru greater than or equal to 110Ag sup (m) > 192Ir > 60Co. (author)

  3. Separation of 90Sr from PUREX HLLW using N,N,N',N'-tetra (2-ethylhexyl)diglycolamide

    This paper describes the separation of 90Sr from PUREX-HLLW employing separation techniques viz. solvent extraction and precipitation. In the first step, PUREX-HLLW was subjected to solvent extraction using TBP (30% in n-dodecane) to remove residual uranium and plutonium. In the subsequent step the raffinate was treated with N,N,N',N'-tetra (2-ethylhexyl) diglycolamide (TEHDGA, 0.20M in 30% isodecyl alcohol and n-dodecane) for the bulk separation of trivalent actinides and lanthanides. The raffinate from this step containing major activity of 90Sr and other fission products such as 137Cs and 106Ru etc. forms ideal feed for 90Sr recovery. Strontium from this non alpha bearing HLLW was extracted using 0.30M TEHDGA in 5% isodecyl alcohol and n-dodecane and stripped with 0.01M HNO3. Recovery of 90Sr was found to be quantitative which was further purified from trace impurities such as 106Ru etc. and concentrated using radiochemical precipitation technique employing Fe scavenging as hydroxide followed by carbonate precipitation after adding natural Fe and Sr as carriers

  4. Screening of silver nanoparticles containing carbonized yeast cells for adsorption of few long-lived active radionuclides

    The present study involves the screening of silver nanoparticles containing carbonized yeast cells isolated from coconut cell sap for efficient adsorption of few long lived radionuclides like 137Cs55, 60Co27, 106Ru44, 239Pu94 and 241Am95. Yeast cells containing silver nanoparticles produced through biological reduction were subjected to carbonization (400 deg C for 1 h) at atmospheric conditions and their properties were analyzed using fourier transform infra-red spectroscopy, X-ray diffraction, scanning electron microscope attached with energy dispersive spectroscopy and transmission electron microscope. The average size of the silver nanoparticles present on the surface of the carbonized silver containing yeast cells (CSY) was 19 ± 9 nm. The carbonized control yeast cells without silver exposure (CCY) did not contain any particles on its surface. The efficiency of CSY and CCY towards the radionuclide adsorption was studied in batch mode at fixed contact time, concentration, and at its native pH. CSY was efficient in removal of 239Pu94 (76.75%) and 106Ru44 (54.73%) whereas CCY showed efficient removal only for 241Am95 (62.89%). Both the adsorbents did not show any retention with respect to 60Co27 and 137Cs55. Based on the experimental data, decontamination factor and distribution coefficient (Kd) were calculated and, from the values, it was observed that these adsorbents have greater potential to adsorb radionuclides. (author)

  5. Monte Carlo-based Bragg-Gray tissue-to-air mass-collision-stopping power ratios for ISO beta sources

    Quantity of interest in external beta radiation protection is the absorbed dose rate to tissue at a depth of 7 mg/cm2 Dt (7 mg/cm2) in a 4-element ICRU (International Commission for Radiation Units and Measurements) unit density tissue phantom. ISO (International Organization for Standardization) 6980-2 provides guidelines to establish this quantity for beta emitters using an extrapolation chamber as a primary standard. ISO 6980-1 proposes two series of beta reference radiation fields, namely, series 1 and series 2. Series 1 covers 90Sr/90Y, 85Kr, 204Tl and 147Pm sources used with beam flattening filter and Series 2 covers 14C and 106Ru/106Rh sources used with beam flattening filter. Dt (7 mg/cm2) is realized based on measured current and set of corrections including Bragg-Gray tissue-to-air mass-stopping power ratio, (S/ρ)t,a. ISO provides (S/ρ)t,a values which are based on approximate methods. The present study is aimed at calculating (S/ρ)t,a for 90Sr/90Y, 85Kr, 106Ru/106Rh and 147Pm sources using the Monte Carlo (MC) methods and compare the same against the ISO values. By definition, (S/ρ)t,a should be independent of cavity length of the chamber which was verified in the work

  6. Monte Carlo-based Spencer-Attix and Bragg-Gray tissue-to-air stopping power ratios for ISO beta sources

    Spencer-Attix (SA) and Bragg-Gray (BG) mass-collision-stopping-power ratios of tissue-to-air are calculated using a modified version of EGSnrc-based SPRRZnrc user-code for the International Organization for Standardization (ISO) beta sources such as 147Pm, 85Kr, 90Sr/90Y and 106Ru/106Rh. The ratios are calculated at 5 and 70 μm depths along the central axis of the unit density ICRU-4-element tissue phantom as a function of air-cavity lengths of the extrapolation chamber l = 0.025-0.25 cm. The study shows that the BG values are independent of l and agree well with the ISO-reported values for the above sources. The overall variation in the SA values is ∼0.3 % for all the investigated sources, when l is varied from 0.025 to 0.25 cm. As energy of the beta increases the SA stopping-power ratio for a given cavity length decreases. For example, SA values of 147Pm are higher by ∼2 % when compared with the corresponding values of 106Ru/106Rh source. SA stopping-power ratios are higher than the BG stopping-power ratios and the degree of variation depends on type of source and the value of l. For example, the difference is up to 0.7 % at l = 0.025 cm for the 90Sr/90Y source. (authors)

  7. Spatial trends on an ungrazed West Cumbrian saltmarsh of surface contamination by selected radionuclides over a 25 year period.

    Caborn, Jane A; Howard, Brenda J; Blowers, Paul; Wright, Simon M

    2016-01-01

    Long term spatial and temporal variations in radionuclide activity have been measured in a contaminated ungrazed saltmarsh near Ravenglass, Cumbria. Over a twenty-five year period there has been a decrease in activity concentration with (106)Ru and (137)Cs showing the highest rate of change followed by Pu alpha and (241)Am. A number of factors contribute to the reduction with time; including radiological half lives, discharge and remobilisation. For (241)Am the lower reduction rate is partially due to ingrowth from (241)Pu and partially as a result of transport of sediment from the offshore Irish Sea mud patch. Considerable spatial variation for the different radionuclides was observed, which with time became less defined. The highest activity concentrations of long-lived radionuclides were in low energy areas, typically where higher rates of sedimentation and vegetation occurred. The trend was reversed for the shorter lived radionuclide, (106)Ru, with higher activity concentrations observed in high energy areas where there was frequent tidal inundation. Surface scrape samples provide a pragmatic, practical method of measuring sediment contamination over large areas and is a sampling approach adopted by most routine environmental monitoring programs, but it does not allow for interpretation of the effect of variation in sedimentation rates. This paper proposes a method for calculating indicative sedimentation rates across the saltmarsh using surface scrape data, which produces results consistent with values experimentally obtained. PMID:26440699

  8. Radioactive contamination of copper produced using nuclear explosives

    Laboratory tests simulating the processing of copper ore after fracturing with nuclear explosives indicate that only very small fractions of the radioactive fission products will be dissolved on leaching with dilute sulfuric acid. Tritium (as tritiated water) will be by far the dominant radionuclide in the circulating leach liquor, assuming use of a fusion device. Only 106Ru appears of significant importance with respect to contamination of the cement copper. It is rejected effectively in electrolytic purification and, therefore, the final copper product should be very low in radiocontamination and not hazardous to the customer. The activity level may be high enough, however, to make the copper unsuitable for some specific uses. If necessary, solvent extraction can be used as an alternative to the cementation process to reduce the radioactivity of the copper products. The tritium in the circulating liquor and the 106Ru in the cement copper are potential hazards at the plant site and must be given consideration in designing and operating the facility. However since the activity levels will be low, the protection necessary to ensure safety of the operating personnel should be neither difficult nor costly to provide. (author)

  9. Studies of environmental radioactivity in Cumbria

    Five stations collecting samples of atmospheric deposition were set up in north Cumbria along a line running inland from the coast for about 17 km. Sampling was continuous from September 1980 to September 1981. Monthly samples were analysed for 106Ru, 137Cs, 144Ce, 238Pu, sup(239,240)Pu, 241Am, 7Be and stable Na, Cl and Al. The objective of the work was to measure the deposition of radionuclides as a function of distance from the sea. By estimating the contributions to the deposition of nuclear weapon test material and of the atmospheric discharges from the British Nuclear Fuels plc works at Sellafield, the effects of the transfer to air and land of radionuclides in the sea could be established. The marine radionuclides were due to the discharges to sea from the Sellafield works. The measurements showed that the deposition was largely due to the sea-to-land transfer process. The highest depositions observed were at 20 m from high water mark, the annual values (rounded, in Bq m-2) being 106Ru, 500; 137Cs, 650; plutonium, 70; 241Am, 30. The highest concentrations in rainwater for the radionuclides studied were less than 3 per cent of the fresh water limits (drinking only) GDL values. The highest estimated accumulations in soil due to atmospheric deposition were less than 1 per cent of the limits. (author)

  10. Release behavior of metallic fission products from pyrocarbon-coated uranium-dioxide particles at extremely high temperatures

    Hayashi, Kimio; Fukuda, Kousaku (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1990-04-01

    Uranium-dioxide particles coated by pyrocarbon (BISO), which were irradiated at 1,300 {approx} 1,400degC to burnups of ca. 1% FIMA, were heated isochronally and isothermally at temperatures between 1,600 and 2,300degC. Release fractions of {sup 137}Cs, {sup 155}Eu and {sup 106}Ru were larger than 10{sup -2} after heating at 2,000degC for 2 h; the results were in contrast to much smaller release fractions from TRISO particles with intact silicon-carbide (SiC) coating. The release of {sup 137}Cs and {sup 144}Ce from the BISO particle was controlled by diffusion in the dense pyrocarbon layer at temperatures between 1,600 and 2,300degC, while that of {sup 155}Eu and {sup 106}Ru was controlled by diffusion in the fuel kernel above 1,800degC. These results can be used as reference data on release behavior of the fission products from TRISO particles with defective SiC layers. (author).

  11. Release behavior of metallic fission products from pyrocarbon-coated uranium-dioxide particles at extremely high temperatures

    Uranium-dioxide particles coated by pyrocarbon (BISO), which were irradiated at 1,300 ∼ 1,400degC to burnups of ca. 1% FIMA, were heated isochronally and isothermally at temperatures between 1,600 and 2,300degC. Release fractions of 137Cs, 155Eu and 106Ru were larger than 10-2 after heating at 2,000degC for 2 h; the results were in contrast to much smaller release fractions from TRISO particles with intact silicon-carbide (SiC) coating. The release of 137Cs and 144Ce from the BISO particle was controlled by diffusion in the dense pyrocarbon layer at temperatures between 1,600 and 2,300degC, while that of 155Eu and 106Ru was controlled by diffusion in the fuel kernel above 1,800degC. These results can be used as reference data on release behavior of the fission products from TRISO particles with defective SiC layers. (author)

  12. The design and the dosimetry of bi-nuclide radioactive ophthalmic applicators

    A novel type of applicator for the treatment of intra-ocular tumors has been developed, based on the two radionuclides 106Ru/106Rh and 125I. The dose distribution of this ophthalmic plaque combines advantageous features of both radionuclides and can be optimally adapted to a tumor thickness in the range 6.5-9 mm, a size which is beyond the dosimetric limitations of the 106Ru/106Rh plaque therapy. Compared with 125I plaques a bi-nuclide plaque allows to maintain the tumor dosage while the dose in the irradiated volume outside of the target volume is significantly reduced. Consequently, radiosensitive structures within the eye can be spared more effectively. Dedicated methods have been developed for the dosimetry of this plaque. These methods are based on our own extensive dosimetric investigations with plastic scintillators. The precondition was the availability, developed in recent years, of a more accurate determination of the absolute dose rate to water of beta- and low energy emitters

  13. Dosimetry of beta-ray ophthalmic applicators: Comparison of different measurement methods

    An international intercomparison of the dosimetry of three beta particle emitting ophthalmic applicators was performed, which involved measurements with radiochromic film, thermoluminescence dosimeters (TLDs), alanine pellets, plastic scintillators, extrapolation ionization chambers, a small fixed-volume ionization chambers, a diode detector and a diamond detector. The sources studied were planar applicators of 90Sr-90Y and 106Ru-106Rh, and a concave applicator of 106Ru-106Rh. Comparisons were made of absolute dosimetry determined at 1 mm from the source surface in water or water-equivalent plastic, and relative dosimetry along and perpendicular to the source axes. The results of the intercomparison indicate that the various methods yield consistent absolute dosimetry results at the level of 10%-14% (one standard deviation) depending on the source. For relative dosimetry along the source axis at depths of 5 mm or less, the agreement was 3%-9% (one standard deviation) depending on the source and the depth. Crucial to the proper interpretation of the measurement results is an accurate knowledge of the detector geometry, i.e., sensitive volume and amount of insensitive covering material. From the results of these measurements, functions which describe the relative dose rate along and perpendicular to the source axes are suggested

  14. Nuclear accident impact on the ecological environment

    This article reviewed the eco-environmental behavior of radionuclides released into the environment by nuclear explosion and nuclear accidents, especially of several key radionuclides with biological significance, including 137Cs, 95Zr, 90Sr, 131I, 3H and 14C, in order to correctly understand the case of nuclear accidents and its pollution, maintain the social stable, and provide suitable measures for environmental protection and safety. (author)

  15. Separation of cesium from other fission products using ZrP-80-APW-20

    An ion exchange method for separation of Cs(137Cs) in the presence of long-lived fission products such as 95Zr, 95Nb etc., was developed. Selectively 137Cs was observed on the column whereas Zr and Nb which were complexed as oxalates passed through. 137Cs was eluted using 4 M nitric acid and 4 m ammonium nitrate solution. (author). 4 refs., 2 tabs

  16. Investigation of some atmospheric circulation variations using cosmogenic radionuclides

    A relationship between solar activity and the time of occurrence of the seasonal concentration maximum of cosmogenic 7Be in the ground air layer is established. In years of minimal solar activity the peak of concentrations can be observed in spring. With increasing Wolf's numbers this peak of concentrations shifts towards the end of summer. An analogical regularity is characteristic for artificial 137Cs and 95Zr + 95Nb. However, it is expressed much weaker. (author)

  17. Ruthenium-106 brachytherapy for thick uveal melanoma: reappraisal of apex and base dose radiation and dose rate

    Jaberi, Ramin; Sedaghat, Ahad; Azma, Zohreh; Nojomi, Marzieh; Falavarjani, Khalil Ghasemi; Nazari, Hossein

    2016-01-01

    Purpose To evaluate the outcomes of ruthenium-106 (106Ru) brachytherapy in terms of radiation parameters in patients with thick uveal melanomas. Material and methods Medical records of 51 patients with thick (thickness ≥ 7 mm and < 11 mm) uveal melanoma treated with 106Ru brachytherapy during a ten-year period were reviewed. Radiation parameters, tumor regression, best corrected visual acuity (BCVA), and treatment-related complications were assessed. Results Fifty one eyes of 51 consecutive patients including 25 men and 26 women with a mean age of 50.5 ± 15.2 years were enrolled. Patients were followed for 36.1 ± 26.5 months (mean ± SD). Mean radiation dose to tumor apex and to sclera were 71 (± 19.2) Gy and 1269 (± 168.2) Gy. Radiation dose rates to tumor apex and to sclera were 0.37 (± 0.14) Gy/h and 6.44 (± 1.50) Gy/h. Globe preservation was achieved in 82.4%. Preoperative mean tumor thickness of 8.1 (± 0.9) mm decreased to 4.5 (± 1.6) mm, 3.4 (± 1.4) mm, and 3.0 (± 1.46) mm at 12, 24, and 48 months after brachytherapy (p = 0.03). Four eyes that did not show regression after 6 months of brachytherapy were enucleated. Secondary enucleation was performed in 5 eyes because of tumor recurrence or neovascular glaucoma. Tumor recurrence was evident in 6 (11.8%) patients. Mean Log MAR (magnification requirement) visual acuity declined from 0.75 (± 0.63) to 0.94 (± 0.5) (p = 0.04). Best corrected visual acuity of 20/200 or worse was recorded in 37% of the patients at the time of diagnosis and 61.7% of the patients at last exam (p = 0.04). Non-proliferative and proliferative radiation-induced retinopathy was observed in 20 and 7 eyes. Conclusions Thick uveal melanomas are amenable to 106Ru brachytherapy with less than recommended apex radiation dose and dose rates. PMID:26985199

  18. Conversion probabilities of low-energy (ℎω≤3 keV) nuclear transitions in the electron shells of free atoms. Article translated from Journal Yadernye Konstanty (Nuclear Constants). Series: Nuclear Constants, Issue No. 1, 1987

    Conversion of some low-energy transitions (ℎω≤3 keV) in the nuclei 90Nb, 99Tc, 103Ru, 110Ag, 140Pr, 142Pr, 153Gd, 159Gd, 160Tb, 165Tm, 171Lu, 173W, 188Re, 193Pt, 201Hg, 205Pb, 236Pa and 250Bk are investigated for the case of an isolated atom. The conversion transition probabilities are calculated using the electron wave functions, obtained through numerical integration of the Dirac equations in the atomic field within the framework of the Hartree-Fock-Slater method. The calculation is carried out for the normal configuration of the valence band of the aforementioned atoms. The calculation results are tabulated in this paper. (author)

  19. Solvent extraction using tetracycline as complexing agent Pt. 14

    The behaviour of tetracycline as an extracting agent for Sr, I, Ba, Mo, Tc, Zr, Nb, Cs, Ru, Te and U was studied and the influence of the acidity of the aqueous phase upon extraction of the elements mentioned was examined. Experiments were made to determine whether the species extracted into the organic phase is the complex formed between tetracycline and the elements considered and to determine the time of shaking necessary so that the equilibrium between the phases is attained. As a practical application, the possibility of using the tetracycline-benzyl alcohol system for separating the fission products sup(137)Cs, sup(140)La, sup(141)Ce, sup(103)Ru, sup(95)Nb from each other and from uranium is presented. The same study was made for sup(131)I, sup(99m)Tc, sup(99)Mo, sup(132)Te, sup(239)Np and uranium and the steps necassary for the separation of these elements are proposed. (author)

  20. Neutron-spectroscopic strength in Ru isotopes

    Duarte, J.L.M.; Borello-Lewin, T.; Horodynski-Matsushigue, L.B. (Instituto de Fisica da Univrsidade de Sao Paulo, Sao Paulo (Brazil))

    1994-08-01

    A systematic, high resolution (6--8 keV) study of ([ital d],[ital t]) reactions on [sup 100,102,104]Ru is reported. Spectroscopic factors were extracted by comparison of experimental angular distributions with distorted wave Born approximation predictions. All of the information for [sup 99]Ru and, for excitation energies above 0.9 MeV, for [sup 103]Ru is new. Most of the strength expected for the 50--82 neutron shell was found. The strength distributions are discussed, also in comparison with the corresponding stripping reactions. Special attention is focused on extremely low and relatively intense [ital l]=3 excitations and on the [ital l]=4 transfer pattern observed.

  1. Multi-tracer study on in vivo and in vitro binding of trace elements with mouse liver DNA

    In vivo, semi-in vivo and in vitro binding of a series of trace elements (Be, Sc, Mn, Co, Zn, As, Se, Rb, Sr, Y, Zr, Tc, Ru and Rh) is studied by the multi-tracer technique. The corresponding nuclides in the multi-tracer solution used are 7Be, 46Sc, 54Mn, 58Co, 65Zn, 74As, 75Se, 83Rb, 85Sr, 88Y, 88Zr, 95Tcm, 103Ru and 102Rhm. It is found that most elements bound mouse liver DNA in vivo except As, Ru and Rh. In the semi-in vivo experiment, only elements Rh and As are not observed to be bound with DNA. In the in vitro experiment, DNA bound with all elements, among which Rb, Se, Zr, Ru and As showed very slight binding. In comparison, the binding in vitro is the strongest, semi-in vivo the medium and in vivo the weakest

  2. The body contents of gamma emitters in adults after the Chernobyl accident and an estimation of exposure for intakes in 1986

    The measuring equipment parameters as well as the results measurement and processing methods applied for estimating the individual body contents of the photon emitters released during the nuclear reactor accident at Chernobyl are presented. The results of estimations of 131I contents in the thyroid and of 103Ru, 136Ru, 134Cs and 137Cs contents in the whole body, based on the measurements performed till the end of 1986 are given. It was observed that the body contents of both the caesium isotopes had been increasing till the end of 1986, which is indicative of prolonged intakes. The average and maximum levels of the committed doses in the thyroid due to 131I uptakes as well as the doses absorbed by the whole body as a result of 134Cs and 137Cs intakes in 1986 have been estimated for the Warsaw region population. 5 refs., 8 figs., 8 tabs. (author)

  3. Behavior of ruthenium, cesium and antimony during simulated HLLW vitrification

    The behavior of ruthenium, cesium, and antimony during the vitrification of simulated high-level radioactive liquid wastes (HLLW) in a liquid fed melter was studied on a laboratory scale and on a semi-pilot scale. In the laboratory melter of a 2.5 kg capacity, a series of tests with the simulate traced with 103Ru, 134Cs and 124Sb, has shown that the Ru and Cs losses to the melter effluent are generally higher than 10% whereas the antimony losses remain lower than 0.4%. A wet purification system comprising in series, a dust scrubber, a condenser, an ejector venturi and an NOx washing column retains most of the activity present in the off-gas so that the release fractions for Ru at the absolute filter inlet ranges between 5.10-3 to 5.10-5% of the Ru fed, for Cs the corresponding release fraction ranges between 3.10-3 to 10-4% and for Sb the release fraction ranges between 1.7 10-4 to 1.7 10-5%. The same experiments were performed at a throughput of 1 to 2 1 h-1 of simulated solution in the semi-pilot scale unit RUFUS. The RUFUS unit comprises a glass melter with a 50 kg molten glass capacity and the wet purification train comprises in series a dust scrubber, a condenser, an ejector venturi and an NOx washing column. The tracer tests were restricted to 103Ru and 134Cs since the laboratory tests had shown that the antimony losses were very low. The results of the tests are presented

  4. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

  5. Recent studies related to head-end fuel processing at the Hanford PUREX plant

    Swanson, J.L.

    1988-08-01

    This report presents the results of studies addressing several problems in the head-end processing (decladding, metathesis, and core dissolution) of N Reactor fuel elements in the Hanford PUREX plant. These studies were conducted over 2 years: FY 1986 and FY 1987. The studies were divided into three major areas: 1) differences in head-end behavior of fuels having different histories, 2) suppression of /sup 106/Ru volatilization when the ammonia scrubber solution resulting from decladding is decontaminated by distillation prior to being discharged, and 3) suitability of flocculating agents for lowering the amount of transuranic (TRU) element-containing solids that accompany the decladding solution to waste. 16 refs., 43 figs.

  6. Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy; Aplicacion del codigo Monte Carlo Penelope a diversos problemas dosimetricos en radioterapia

    Sanchez, R.A.; Fernandez V, J.M.; Salvat, F. [Servicio de Oncologia Radioterapica. Hospital Clinico de Barcelona. Villarroel 170 08036 Barcelona (Spain)

    1998-12-31

    In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: {sup 192} Ir, {sup 125} I, {sup 106} Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)

  7. Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy

    In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: 192 Ir, 125 I, 106 Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)

  8. The hot bench scale plant Ester for the vitrification of high level wastes

    In this paper the hot bench-scale plant ESTER for the vitrification of the high-level radioactive wastes is described, and the main results of the first radioactive campaign are reported. The ESTER plant, which is placed in the ADECO-ESSOR hot cells of the C.C.R.-EURATOM-ISPRA, has been built and is operated by the ENEA, Departement of Fuel Cycle. It began operating with real radioactive wastes about 1 year ago, solidifying a total of 12 Ci of fission products into 2,02 Kg of borosilicate glass, corresponding to 757 ml of glass. During the vitrification many samples of liquid and gaseous streams have been taken and analyzed. A radioactivity balance in the plant has been calculated, as well as a mass balance of nitrates and of the 137Cs and 106Ru volatized in the process

  9. The Chernobyl fallout in Greece and its effects on the dating of archaeological materials

    The effects of the fallout from the nuclear reactor accident at Chernobyl have been monitored at various sites in Greece. Here we present the first estimates of gamma dose rates, an essential parameter in the dating of archaeological materials by thermoluminescence (TL) and ESR methods. The dose rates are derived from the long-lived radionuclides of 137Cs, 134Cs, 106Ru and 144Ce (with t1/2 ≥ 1 yr). The present dose rates vary between 30 and 60 mrad/yr, but maximum values of around 811 mrad/yr have also been recorded, for ground-surface exposures. These dose rate values must be regarded as very significant to TL and ESR dating of samples from now on and a correction factor should be applied. (orig.)

  10. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137Cs, 106Ru, 148Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  11. Cerenkov counting as a complement to liquid scintillation counting

    A commercially available spectrometer was calibrated for liquid scintillation (LS) and Cerenkov counting efficiency (CCE) using National Institute of Standards and Technology traceable solutions. The CCE increased linearly over a 3 order of magnitude range in 40K β activity, and by 42% per MeV as β-energies increased from 0.300 to 3.54 MeV, achieving a maximum value of 80% for 106Ru/106Rh The CCE can be enhanced by 10-15% when a wavelength shifter is used. A comparison of the data showed that the CCE was typically 20-50% less than the LS counting efficiency for β-particles with maximum energies >1 MeV. Applications that utilize sequential CCE and LS counting to quantitate activity concentrations are discussed for samples containing two β-emitting nuclides of differing energies. (Author)

  12. Ruthenium nitrosyl complexes in nitric acid solutions

    Nine nitrosyl ruthenium complexes have been separated and identified in aqueous solutions of nitric acid. The separation method was low temperature, gradient elution, reverse phase partition chromatography using tri-n-butyl phosphate on a kiesel gel 60 support using 106Ru labelled complexes in the nitric acid phase. The identification of the complexes was deduced from the relationships between the products of aquation and nitration and paper chromatography using both methyl-iso-propyl ketone and nitric acid-acetone elutions. The proportion of each complex at equilibrium in various concentrations of nitric acid have been measured. The rates of nitration in 10 M nitric acid, and of aquation in 0.45 M nitric acid have been determined at 00C. (author)

  13. Radioecological Impact of the French Nuclear Power Plants on the Marine Environment

    Since the end of the 1970s a global method has been developed and improved to characterise the radioecological impact of French nuclear power plants (PWRs) on the marine ecosystems. The environment of every nuclear plant is examined yearly, in addition to special studies carried out before plant operations begin and after each period of ten years. Three nuclear power plants are situated on the Channel coast (Flamanville, Paluel, Penly) and one on the North Sea coast (Gravelines). Near the power plants local radioecological impact is measurable and is essentially due to 58Co, 60Co and 110Agm. The monitoring of artificial gamma-emitter radioactivity in bioindicators (Fucus sp.) reveals the overall decline in releases from the four power stations in question (58Co, 60Co and 110Agm) as well as the more marked decrease in relation to the reprocessing plant at La Hague (106Ru, 60Co, 125Sb, 241Am). (author)

  14. Participation of the radioanalytical laboratories of CDTN/CNEN in the national intercomparison program for radionuclide analysis in water

    This paper reports the performance of the radioanalytical laboratories of Analytical Techniques Service - SERTA/CDTN/CNEN achieved in the National Intercomparison Program - PNI conducted by the Institute of Radiation Protection and Dosimetry - IRD. The program comprehends the distribution of synthetic water samples for some (approximately 25) national laboratories in order to promote analytical credibility and to ensure the reliability of the analytical results based on radiochemical methodologies. In this program water samples are artificially contaminated at environmental level with known amounts of important radionuclides for radiological protection. The parameters and analytes evaluated are, 60Co, 65Zn, 106Ru, 133Ba, 134Cs, 137Cs, 3H, 90Sr, 226Ra, 228Ra, 210Pb, Unat, 232Th and gross alpha and beta measurements. In the last 13 years our performance was considered 'good' and 'acceptable' in over 90% of the rounds accomplished. (author)

  15. Chromatographic decontamination of medium-activity waste concentrates

    The chromatographic decontamination of a MAW concentrate was carried out in a laboratory plant in 1-l-batches in the following way: In order to purify the nitric MAW concentrate from its solid and organic contamination products, it is passed through a filter and an absorber (SM7) for organic species. Subsequently the purified solution runs on-line through all following columns. First the main activity carrier cesium (137Cs, 134Cs) is transferred to ammonium molybdate phosphate (AMP-1) by means of a newly developed fluidized bed process. In the further course, 125Sb is separated on metal oxides (Sb2O5, MnO2) and the three-valued actinides/lanthanides on an extraction-chromatographic CMPO column. Finally the remaining 106Ru and 60Co activities are separated on dimethylglyoximes (DMG) coated on active carbon. (orig./RB)

  16. Radiation Protection in Brachytherapy. Report of the SEFM Task Group on Brachytherapy

    This document presents the report of the Brachytherapy Task Group of the Spanish Society of Medical Physics. It is dedicated to the radiation protection aspects involved in brachytherapy. The aim of this work is to include the more relevant aspects related to radiation protection issues that appear in clinical practice, and for the current equipment in Spain. Basically this report focuses on the typical contents associated with high dose rate brachytherapy with 192Ir and 60Co sources, and permanent seed implants with 125I, 103Pd and 131Cs, which are the most current and widespread modalities. Ophthalmic brachytherapy (COMS with 125I, 106Ru, 90Sr) is also included due to its availability in a significant number of spanish hospitals. The purpose of this report is to assist to the medical physicist community in establishing a radiation protection program for brachytherapy procedures, trying to solve some ambiguities in the application of legal requirements and recommendations in clinical practice. (Author)

  17. Game as a bioindicator of the radiocontamination

    Natural and artificially produced radionuclides were determined in meat and bones of deer, boar and wild hare on hunting areas in Vojvodina (Serbia). Seven natural radionuclides and three fission products (235U, 238U, 232Th, 7Be, 144Ce, 40K, 106Ru, 134Cs, 137Cs, 90Sr) were identified in the investigated game samples. The highest contents of the radionuclides were found in bones and meat of boars and the lowes in the bones of fallow-deer. The predominant radionuclides were 40K and 90Sr, for all of the investigated animals and their contents depended strongly upon the game species, organ type and the age of the animal. The examined breeding sites did not appear to have any effect on the radionuclide contents in game, which indicates that the radionuclides were uniformly distributed over the habitat. (author) 5 refs.; 3 tabs

  18. Diffusion coefficients of fission products in the UO sub 2 kernel and pyrocarbon layer of BISO-coated fuel particles at extremely high temperatures

    Hayashi, Kimio; Fukuda, Kousaku (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1990-11-01

    Release of metal fission products from pyrocarbon (PyC) coated UO{sub 2} particles was studied by post-irradiation annealing at temperatures from 1600 to 2300deg C. Release of {sup 106}Ru and {sup 155}Eu was controlled by diffusion in the kernel at temperatures above 1800deg C, and their reduced diffusion coefficients in the kernel were very close to each other. The diffusion coefficient of Cs, D{sub Cs} (m{sup 2}/s), in the PyC layer was determined from the fractional release, as follows: D{sub Cs}=1.2x10{sup -3} exp(-4.12x10{sup 5} (J/mol)/RT), which was larger than that of Ce by an order of magnitude. The diffusion coefficients of fission products in the PyC layer was discussed in terms of their ionic radii and stability of their carbides. (orig.).

  19. Diffusion coefficients of fission products in the UO2 kernel and pyrocarbon layer of BISO-coated fuel particles at extremely high temperatures

    Release of metal fission products from pyrocarbon (PyC) coated UO2 particles was studied by post-irradiation annealing at temperatures from 1600 to 2300deg C. Release of 106Ru and 155Eu was controlled by diffusion in the kernel at temperatures above 1800deg C, and their reduced diffusion coefficients in the kernel were very close to each other. The diffusion coefficient of Cs, DCs (m2/s), in the PyC layer was determined from the fractional release, as follows: DCs=1.2x10-3 exp[-4.12x105 (J/mol)/RT], which was larger than that of Ce by an order of magnitude. The diffusion coefficients of fission products in the PyC layer was discussed in terms of their ionic radii and stability of their carbides. (orig.)

  20. Evaluation of artificial radioactivity of the north Western mediterranean sea and evaluation of the sanitary consequences

    The results of radiological measurements of the north west mediterranean observation network outline the level of artificial radionuclides coming from industrial seewages, 106Ru and from atmospheric fall out, 137Cs and sup(239+240)Pu measured on 3 differents types of bioindicators: Mytilus sp., Posidonia oceanica (L.) Del. and demersal fishes as Solea sp., Anguilla anguilla L., Conger conger L. Mytilus sp. is quite a perfect bioindicator of radionuclides contamination but must be linked with fishes sampling which muscles concentrate Cesium at higher level. The sanitary consequences for the waterside population involved by molluscs and fishes ingestion contamined by these 3 radionuclides lead to a fraction (10-5) of the annual dose limit recommanded by the ICRP 26

  1. Trace elements retained in washed nuclear fuel reprocessing solvents

    Analysis of purified TBP extractant from solvent extraction processes at Savannah River Plant showed several stable elements and several long-lived radioisotopes. Stable elements Al, Na, Br, Ce, Hg, and Sm are found in trace quantities in the solvent. The only stable metallic element consistently found in the solvent was Al, with a concentration which varies from about 30 ppM to about 10 ppM. The halogens Br and Cl appear to be found in the solvent systems as organo halides. Radionuclides found were principally 106Ru, 129I, 3H, 235U, and 239Pu. The 129I concentration was about 1 ppM in the first solvent extraction cycle of each facility. In the other cycles, 129I concentration varied from about 0.1 to 0.5 ppM. Both 129I and 3H appear to be in the organic solvent as a result of exchange with hydrogen

  2. Distribution and migration of radionuclides in soils and their uptake in plants after the Chernobyl reactor accident

    A method of sampling from deep soil layers was developed. With this technique, 500 samples from two sites in Styria, Austria were taken from depth up tp 80 cm and from arable soils as well as from meadows, in the years 1986, 1987 and 1988. The samples were prepared and analyzed for Cs 134, Cs 137, Ru 106, Ru 103, Sb 125, K 40 and Sr 90. The resulting depth distributions are presented in tables and graphs. The transfer functions soil-to-plant (especially wheat) are derived. The conclusion is that cesium remains tightly bound, in meadows, to a surface-near layer for years. This is not the case with Sr 90. 31 refs., 33 tabs., 181 figs

  3. A direct timing method for the two-dimensional precision coordinate detectors based on thin-walled drift tubes

    The results of a study of the longitudinal spatial resolution of 2 m long straw tubes by means of the direct timing method (DTM) are presented. The feasibility of achieving a coordinate resolution (r.m.s.) better than 9 mm over full length of the straw is demonstrated. The spatial resolution insignificantly changes when measured by detecting gammas from a 55Fe gamma-ray source or minimum ionizing particles from a 106Ru source. The use of the same type of FEE for data taking both for measuring the drift time of ionization electrons and propagation of a signal along the anode wire allows one to construct a two-dimensional detector for precision coordinate measurements

  4. Influence of the complexones on the ionic transfer through cell membranes and the level of radionuclides-metals accumulation in water plants

    The influence of the complexones diethylenetriaminepentaacetic acid (DTPA), ethylenediaminetetraacetic acid (EDTA) and hydroxyethylidenediphosphonic acid (HEDP) with relatively low concentrations (from 0.1 to 50 mg/l) on ionic permeability of cell membranes of Nitellopsis obtusa algae, as well as on the accumulation levels of the radionuclides 144Ce, 106Ru, 90Sr and 137Cs in different species of water plants has been studied. It has been shown that complexones under study (with the concentrations up to 50 mg/l) may reduce accumulation levels of three- and four-valent metals and their radionuclides in water plants. For the plants in natural as well as in artificial nutrient medium, however, the complexones increase the availability of metals, forming with the readibly soluble, mobile complex compounds

  5. Critical pathway studies for selected radionuclides. Part of a coordinated programme on environmental monitoring for radiological protection in Asia and the Far East

    The programme carried out critical pathway studies for selected radionuclides (60Co, 63Ni, 59Fe, 54Mn, sup(110m)Ag, 106Ru and 144Ce) and assessed population exposure in the vicinity of Tarapur Atomic Power Station. The following topics are covered under the programme. (i) Demographic study of dietary habits and consumption data for Tarapur population. (ii) Concentration and accumulation of radionuclides in food products. (iii) Determination of radionuclides in sea water, silt, marine algae and marine organisms at Tarapur Atomic Power Station (TAPS) Site. (iv) Behaviour of radionuclides released to marine environment. (v) Evaluation of critical exposure pathway. (vi) Population exposure in the vicinity of Tarapur Atomic Power Station

  6. Radioactivity concentrations in the North Sea

    The following nuclides can be detected in the North Sea: 90Sr, 106Ru, 137Cs, 134Cs, 241Am and 244Cm. The water in the channel east of Cherbourg still contains low quantities up to 0.5 pCi/l of the fission product 125Sb (t 1/2=2.77 a). The activity of T is not exactly known, for the german Bay it is estimated to 50 pCi/l. In comparison of these artificial radioactivies with the natural radioactivity of the North Sea, which mainly depends on 40K, 87Rb as well as uranium and decay products, it has to be taken into account that the activity of α- and β-active substances from nuclear weapons tests before 1962 and by controlled discharge from nuclear power plants, amounts to a range of 1%. (HP)

  7. Importance of colloids in the transport within the dissolved phase (<450 nm) of artificial radionuclides from the Rhone river towards the Gulf of Lions (Mediterranean Sea)

    The significance of colloidal fractions regarding the transport of artificial radionuclides in natural water systems is underlined by using sequential ultrafiltration both in the Rhone freshwater and the marine area under and outside the influence of the river outflow. Indeed, the Rhodanian aquatic system represents an interesting test site as various artificial radionuclides are released into the Rhone river by several nuclear installations. We focused our study on 137Cs, 106Ru, 60Co, 238Pu and 239+240Pu. Our results show that Fe, Al and Organic carbon (OC) are the main components of colloidal matter. Colloids represent about 15% of dissolved (238Pu and 239+240Pu and have no significance on 137Cs flux

  8. Intercomparison of radionuclides measurements in marine cockle flesh sample IAEA-134

    The results of an intercomparison exercise on a cockle flesh sample from Irish Sea, IAEA-134, designed for the determination of artificial and natural radionuclide levels, are reported. The data from 134 laboratories representing 49 countries have been evaluated. The following are the recommended values, with confidence intervals, for 40K, 60Co, 137Cs, and 239+240Pu (Reference date: 1 January 1992). Information values for 90Sr, 106Ru, 125Sb, 134Cs, 154Eu, 155Eu, 210Pb, 210Po, 226Ra, 228Ra, 228Th, 230Th, 232Th, 234U, 235U, 238U, 238Pu and 241Am are also reported. All the following values are expressed in Bq kg-1 (dry weight). (author)

  9. Intercomparison of radionuclide measurements in marine sediment sample IAEA-135

    The results of an intercomparison exercise on a marine sediment from Irish Sea, IAEA-135, designed for the determination of artificial and natural radionuclides levels, are reported. The data from 151 laboratories representing 51 countries have been evaluated. The following are the recommended values, with confidence intervals, for 40K, 60Co, 134Cs, 137Cs, 154Eu, 155Eu, 226Ra, 228Ra, 232Th,238Pu, 239+240Pu (Reference date: 1 January 1992). Information values for 57Co, 90Sr, 106Ru, 125Sb, 210Pb, 210Po, 228Th, 230Th, 234U, 235U, 238U and 241Am are also reported. All values are expressed in Bq kg-1 dry weight. (author)

  10. Batch extraction studies for the recovery of 233U from thoria irradiated in PHWR

    Batch equilibrium studies were carried out to optimise the extraction parameters for the recovery of 233U from thoria irradiated in PHWR. The thorium concentration and the acidity of the feed was adjusted to ca. 100 g/l and 4 M nitric acid respectively. The concentration of uranium was in the range of 1.4 g/L and it contained long lived fission product like 144Ce-144Pr, 134Cs, 137Cs, 106Ru-106Rh, 105Eu, 154Eu, 90Sr-90Y and 125Sb. 3% TBP in dodecane was used as the solvent. Four stages of batch extraction was followed by a single scrub stage of 4 M nitric acid. The scrubbed organic was stripped with 0.01 M HNO3 thrice. The stripped product was concentrated by evaporation and passed through a cation exchanger to remove the residual thorium. The results of the studies are discussed in detail. (author)

  11. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    Wang Tienko E-mail: tkw@faculty.nthu.edu.tw; Peir Jinnjer

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I, and {sup 140}La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, {sup 235}U burn-up values can be deduced by iterative calculations. The complication caused by {sup 239}Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products {sup 137}Cs, {sup 134}Cs/{sup 137}Cs ratio and {sup 106}Ru/{sup 137}Cs ratio.

  12. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio

  13. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

  14. A gauge for measuring the dose rate and activity of ophthalmic applicators

    A gauge is developed for determining the dose rate distribution and surface activity of ophthalmic brachytherapy applicators, particularly 106Ru applicators. A plastic Φ2x2 mm scintillator is used as the radiation detector, featuring a high pulse count rate, which results in a law 0.5% random error, due to good counting statistics. Automatic gain control of the photomultiplier tube (PMT) is achieved using a LED as the reference light source. The PMT operates in pulse mode. Long term gain variation due to fatigue of the PMT or ambient temperature variation is thus compensated for. The count rate error due to inaccurate setting of the high voltage supply of the PMT is 0.4%, and the instability error over 7 hours of continuous operation does not exceed 102%, peak-to-peak. (author)

  15. Characterization of gaseous and particulate effluents from the nuclear vitrification project

    Samples were taken during the second high-level liquid waste vitrification campaign associated with the NWVP. Sample analysis established the following total average emission levels in the undiluted vitrification off-gas stream: 3H - 1.2 μCi/m3, 14C - 0.2 μCi/m3, 129I - 1 nCi/m3, NO/sub x/ - 0.5%, 99Tc - 1.6 pCi/m3, and γ-emitters - 105(γ/s)/m3. The aerosol size distribution is composed almost entirely of particles exhibiting smaller diameters than the minimum value for which absolute filters are rated, with an empirical geometric mean diameter of 0.13 μm and a particle mass concentration of 84.7 pg/cm3. The particulate matter was composed of 106Ru, 125Sb, /sup 125m/Te, 134Cs, 137Cs, 144Ce, 154155Eu and 241Am. The particulate emission levels in the undiluted process off-gas stream were: 106Ru - 44 nCi/m3, 125Sb - 0.52 nCi/m3, /sup 125m/Te - 1.7 nCi/m3, 134Cs - 7.8 nCi/m3, 137Cs - 52 nCi/m3, 144Ce - 0.8 μCi/m3, 154Eu - 0.3 nCi/m3, 155Eu - 0.18 nCi/m3, and 241Am - 0.19 nCi/m3. All these environmental gaseous and airborne emissions level liquid waste were well within PNL waste management guidelines

  16. Cycling of radionuclides and impact of operational releases in the near-shore ecosystem off the west coast of India

    Radioecology of the near-shore environment was investigated at Bombay and Tarapur on the west coast of India. The major radionuclides released from the processing plant at Bombay were 137Cs, 144Ce and 106Ru and were discharged through a pipeline to Bombay Harbour bay. At Tarapur the major radionuclides discharged were 131I, 134Cs, 137Cs and 60Co. After dilution with condenser-coolant sea water, releases were carried out through open channels along the shoreline to the sea. Studies on radionuclide cycling in clam-bed sediment, the indicator type benthic organism Anadara granosa, and the fish gobiid mudskipper in Bombay Harbour bay showed that the effective half-life of 137Cs is short compared with the physical half-life (30 years). This is attributable to the desorption of 137Cs from sediment and the fact that the benthic organism readily equilibrates with its environment. The dose to the benthic organism was calculated to be about 0.06 to 3μGy/h. Desorption of 144Ce and 106Ru was not observed. Investigations at Tarapur showed the effective distribution of radionuclides in sea water, sediment, seaweed and marine organisms. It was observed that the radionuclides discharged were mainly confined to a region 2 km from the outfall. The highest activity found in these matrices was only 7% of the derived maximum permissible concentration. The highest thyroid dose due to 131I by fresh seafood intake was 3% of the permissible dose and the highest whole-body dose due to other radionuclides was only 1% of the permissible dose. (author)

  17. Medical physics aspects of ophthalmic brachytherapy

    Intraocular melanoma is the most common primary malignancy of the eye. Radiation therapy using ophthalmic plaque has proved successful in the management of various ocular lesions. Although a few centres were using 90Sr/90Y plaques for shallow turtlours some years ago, eye plaque therapy was not a common practice in India. A revived interest in the use of eye plaque therapy and very high cost of imported sources has led to the development and production of 125I seed sources by the Radiopharmaceuticals Division, BARC. This report presents a brief description on the clinical, dosimetry and radiation safety aspects of 90Sr/90Y and 106Ru/106Rh beta ray and 125I gamma ray eye plaque applicators. This report has been divided in five Sections. Section I presents general introduction of ophthalmic brachytherapy including the structure of a human eye, types of ophthalmic plaques and characteristics of radioisotopes commonly used in such applications. A brief review of sources, applicators and dosimetry of 90Sr/90Y and 106Ru/106Rh beta and 125I gamma ophthalmic plaques are given in Section II and Section III, respectively. Section IV contains the single seed dosimetry data of BARC OcuProsta 125I seed as well as dosimetry data of typical eye plaques loaded with BARC OcuProsta 125I seed. Quality assurance and radiation safety aspects of these eye applicators are described in Section V. A proforma of the application required to be filled in by the user institution for obtaining regulatory consent to start eye plaque therapy has also been appended to this report. (author)

  18. The Influence of Body Size and Food Preparation Practices on the Uptake and Loss of Radionuclides in Cumbrian Winkles

    It has been suggested that larger winkles may show higher concentrations of radioactivity and that critical group members, when providing MAFF with sub-samples of winkles they have collected, retain the largest organisms for consumption. If both these hypotheses are true, this could imply higher doses than typically estimated. Conversely, it is assumed that winkles are consumed immediately following collection, without gut clearance, which may overestimate the consumption dose. Results obtained in this study indicate the following. (1) Inverse correlations are observed between concentration and body size for 106Ru and 241Am. A positive relationship is observed for 60Co. Although concentrations of 137Cs decrease with increasing body size the negative correlation is not significant, while 110Agm concentration is independent of body size. (2) There is no evidence that critical group consumers retain larger organisms. In any case, consuming larger or smaller winkles, over the size range tested, probably does not affect dose uptake by more than 10%. (3) Concentrations of plutonium and americium decline to 50% of original by soaking winkles for 18 h. Concentrations of caesium are also reduced. (4) Concentrations of 106Ru, 110Agm and 60Co are not reduced by soaking. Overall, current approaches to deriving consumption doses are not likely to suffer from selection bias. Overnight soaking of winkles in saline solution could decrease radiation doses to consumers by a factor of nearly 2; provided that the fractional gut transfer of activity remaining is unaffected. This last assumption, however is questionable and the current approach should provide a 'reasonably conservative' dose estimate appropriate to critical group studies. (author)

  19. 1971 Animal Investigation Program. Annual report

    Data are presented that were obtained from the radioanalysis of tissues collected from cattle, deer, desert bighorn sheep, and other wildlife that reside on or near the Nevada Test Site. Cesium-137 and 106Ru were the only gamma-emitting radionuclides detected in the soft tissues of range cattle. Ruthenium-106 was detected only in the lungs of animals sampled in May. Strontium-90 levels in the cattle femurs ranged from 2 to 37 pCi/g of ash. The latter value was found in the bones of a 14-year-old cow that had lived on the Nevada Test Site her entire life. The bones of the same animal also had the highest level of 239Pu (46 pCi/g of ash) that was reported. Analysis of her 8-month-old fetus revealed the presence of detectable levels of 239Pu which indicates placental transfer of this radionuclide. The average 90Sr levels in the bones from deer and desert bighorn sheep were 3.2 and 4.7 pCi/g of ash, respectively. Elevated levels of 106Ru and 3H were found in the tissues of two mule deer collected near the drainage ponds that collect runoff waters from mines used for nuclear testing activities. Other animals sampled included Golden eagles, feral horses, coyotes, and chukar. The 137Cs levels in an eagle collected during 1964 varied only slightly from one collected during 1971. No gross or microscopic lesions were detected that could be attributable to the effects of ionizing radiation. (auth)

  20. Radioecological impact of the Chernobyl accident on continental aquatic ecosystems

    The pooling of knowledge on water, sediments, aquatic plants and fish allowed an evaluation report to be drawn up on the impact of Chernobyl accident and to extract data on the mechanisms in the transfer of certain radionuclides in rivers and lakes. The radioactivity is related to the level of deposits, essentially, in wet form. Differences in radioactivity levels are noted owing to the distance from Chernobyl, the atmospheric streams and pluviometric conditions. The most commonly detected radionuclides are: 131I, 132Te, 134+137Cs, 103+106Ru, 110m Ag and, to a lesser degree, 89Sr and 90Sr. Very quickly, 137Cs becomes dominant. The peak of radioactivity in rivers occurred very soon after the accident. It was of short duration and the decrease in radioactivity was very quick due to dilution. In lakes, this decay was much slower. In sediment, the radioactivity varied in time owing either to new deposits or to the migration of those deposits downstream in the river basins. The radionuclides present in fallout can be quickly detected using aquatic plant. In certain areas, the concentration of 137Cs increased 200-fold in a few hours. In fish, the presence of 134+137Cs, 103+106Ru, 110m Ag and 90Sr are noted. The only radionuclide of which fixing dynamics can be followed is 137Cs. River fish was only subjected to water and food with a high radioactivity for a very short time and their 137Cs concentration remained constantly low. The effective half-life of 137Cs observed in situ for fish is from 100 to 200 days. For lacustrine fish, we observe differences in radiocontamination, according to the regions (from 48,000 Bq.kg-1 w.w., in Sweden, to 110 in the North of Corsica or the Netherlands), in lakes (in Northern Italy, 137Cs concentrations in fish are higher in small lakes), and species

  1. Radiogenic late effects in the eye after therapeutic application of beta radiation

    Beta irradiation with 90Sr/90Y is used to treat epibulbar tumours (carcinoma, melanoma) and irradiation with 106Ru/106Rh is used to treat intra-ocular tumours (melanoma, retinoblastoma). Two studies have been carried out. Since 1960, 185 patients with epibulbar pigment tumours and 15 patients with conjunctiva carcinomas have been treated with 90Sr/90Y-applicators and observed for several years. The dose applied was 10,000 to 20,000 rads at the focus depending on the type and extent of the tumour. Apart from teleangiectasias of the conjunctiva, there were only a few cases of severe radio-induced complications such as keratopathies and secondary glaucoma, which were regarded as the lesser evil in comparison with the main disease. The radiation cataract after beta irradiation remains peripheral and does not impair vision. So far 39 patients with choroid melanomas and 22 children with retinoblastomas have been observed for more than 5 years after beta irradiation with 106Ru/106Rh. The dose applied at the sclera surface was 40,000 to 100,000 rads for 4 to 8 days. In 39 patients with successfully irradiated choroid melanomas, radio-induced late complications developed such as macula degeneration, opticus atrophy and retinal-vessel ablations, which may impair vision. In the 22 children irradiated, only 7 cases of late complications with impaired functions could be observed. Whereas radiation-induced late damage after beta irradiation of the front section of the eye is of small clinical importance, especially in older patients, intra-ocular tumours with radio-induced late damage in the retinal vessel and capillary system have to be expected after high-dose beta irradiation

  2. Retrospective dosimetry of an accidental intake case of radioruthenium-106 at the Tokai reprocessing plant

    On November 30, 1978, two workers in the acid recovery cell of the nuclear spent fuel reprocessing plant of JAEA-NFCEL were involved in an accident in which they became unconscious due to lack of oxygen. They were rescued immediately by other workers and were given artificial respiration to restore their normal breathing. Subsequent measurements by the whole-body counter showed that they were contaminated internally with 106Ru. Prolonged lung monitoring was carried out for one of them. A significantly high activity of 106Ru was obtained in the lung monitoring on the day of the accident. The physicochemical characteristics of the incorporated radioactive materials were not observed. In order to perform more reasonable internal dose assessment, the interpretation of the bioassay datasets of the worker was made based on the guideline demonstrated in the EU project IDEAS. The effective half-life of the materials in the lungs was determined to be 140 days which leads to the default Type S absorption type in the HRTM and the f1 value was estimated to be less than 0.005 which is one-tenth of the default value. Simultaneous intakes via inhalation and ingestion were also suggested from several pieces of evidence although pure inhalation was assumed for internal dose assessment at the time of the accident. The aerosol size of the materials was not determined due to a lack of information if assuming simultaneous intakes; however, the resulting committed effective dose was about 1 mSv and its variation was small against the aerosol size ranging from 1 μm to 20 μm. (author)

  3. The accumulation of radionuclides by Dreissena polymorpha molluscs - The Accumulation of Radionuclides from Water and Food in the Dreissena polymorpha Mollusks

    Jefanova, O.; Marciulioniene, D. [Nature Research Centre, Akademijos str. 2, LT-08412 Vilnius, Lietuva (Lithuania)

    2014-07-01

    The specific activity of {sup 137}Cs, {sup 60}Co, {sup 54}Mn, {sup 90}Sr was measured in mollusks Dreissena polymorpha samples from lake Drukshiai that is the cooling pond of the Ignalina NPP. Item the accumulation of {sup 137}Cs, {sup 90}Sr, {sup 144}Ce, {sup 106}Ru in the mollusks from water and from phytoplankton which is a part of their diet was evaluated under laboratory conditions. The data of long-term (1991-2009) studies conducted at six monitoring stations of lake Drukshiai show that in 1991 {sup 137}Cs in mollusks Dreissena polymorpha was found only in that lake's area which was influenced by the effluent that got into lake from the industrial drainage channel of Ignalina NPP. In later periods of the investigation the {sup 137}Cs specific activity was detected in mollusks samples which had been collected at other monitoring stations (the aquatory of lake Drukshiai). Meanwhile {sup 60}Co and {sup 54}Mn in Dreissena polymorpha were detected only in that lake's area which was impacted by the industrial drainage channel. The data of long-term investigation show that the major amount of radionuclides has come into lake Drukshiai through the industrial drainage system of Ignalina NPP. Albeit {sup 137}Cs, {sup 90}Sr, {sup 144}Ce, {sup 106}Ru get into the mollusks through a large amount of the water rather than from the food (phytoplankton), therefore the food can also be the main source of radionuclides in the organism of these mollusks in aquatic environment when there are low levels of specific activities of these radionuclides. (authors)

  4. Avian radioecology on a nuclear power station site. Technical progress report, 1 July 1974--30 June 1975

    The continuation of a program demonstrating that free-ranging wild birds can be used to assess environmental radionuclide levels is described. Wild passerine birds are trapped at a nuclear power station site and at two control sites, uniquely marked, non-destructively counted for levels of gamma-emitting radionuclides, and released. Subsequent recapture rates are as high as 80 percent for certain species. Nuclides detected included 40K, 95Zr-95Nb, 137Cs, and apparent 131I, the latter at levels just above detection limits (0.07 pCi/g). Significant variations in mean 137Cs body burdens in Blue Jays and Bobwhite have been observed between sites less than 6 km apart. A significant temporal decrease in 137Cs body burdens has been observed in various species of birds only at the reactor site. Vegetation and meteorological studies have been initiated to help explain these body-burden differences. The effective half-life of 137Cs in the Blue Jay is 6.7 +- 1.5 days. The highest observed level of 95Zr-95Nb (1.08 +- 0.07 pCi/g) was in a Grey Catbird. The effective half-life of the 95Zr-95Nb was 69.9 +- 15 days, corresponding to the physical half-life of this isotope pair (65.5d), and the activity abruptly disappeared after the twelfth day of captivity suggesting that it was present in particulate form either on the feathers or skin, or in the bird's pulmonary system. Inter-laboratory comparisons of radionuclide measurements are reported. The number of birds banded during the current reporting period is 2720, while 2047 specimens were radioassayed, an increase of 9 percent and 313 percent, respectively. An exercise to test the practicality of obtaining avian samples from remote sites (i.e., > 50 miles away) for radionuclide measurement was performed satisfactorily. (U.S.)

  5. The dispersion of pollutants in the Romanian rivers Olt, Somes and Danube

    The paper describes the results obtained in three tracer experiments carried out into the Romanian rivers Olt and Somes as well as into the Romanian part of the Danube by using the radioactive tracer 82Br. The possibility of a radioactive discharge to occur from the Cernavoda Nuclear Power Plant bring to focus also knowing of the distribution of fission products between water and solid suspended material. Consequently there were also studied and presented the distribution coefficients of the radionuclides 131I, 99Mo, 144Ce, 137Cs, and 95Zr for clay, loess and the particular alluvium collected from the Danube bed. (author)

  6. Estimation of thermal neutron flux from natZr activity

    Neutron transmutation doped (NTD) Ge thermistors are developed as low temperature thermometry (in mK range) in the cryogenic Tin bolometer, the India-based TIN detector (TIN.TIN). For this purpose, semiconductor grade Ge wafers are irradiated with thermal neutron at Dhruva reactor, BARC and dopant concentration critically depends on thermal neutron fluence. In order to obtain an independent estimate of the thermal neutron flux, natZr is used in one of the irradiations. The irradiated natZr samples have been studied in the Tifr Low background Experimental Setup (TiLES). The thermal neutron flux is estimated from the activity of 95Zr

  7. Appendix to Health and Safety Laboratory environmental quarterly report. [Fallout radionuclides deposited and in surface air at various world sites; /sup 137/Cs and /sup 90/Sr in milk and drinking water in New York City; and stable Pb in surface air

    Hardy, E.P. Jr.

    1977-07-01

    Tabulated data are presented on the deposition of fallout /sup 89/Sr and /sup 90/Sr at various world land sites through 1976; the ..gamma.. spectra and content of /sup 7/Be, /sup 95/Zr, /sup 137/Cs, /sup 144/Ce, /sup 90/Sr, /sup 210/Pb, /sup 238/Pu, /sup 239/Pu, and stable Pb in samples of surface air collected during 1966 at various world sites; and the content of fallout /sup 137/Cs and /sup 90/Sr in samples of drinking water and milk collected in New York City through 1976. (CH)

  8. Bio-accumulation kinetics of radioruthenium in marine bivalves-laboratory study

    Bio-accumulation in marine organisms is a basic and important interaction for the retention and migration of pollutants such as heavy metals and radioactive nuclides in marine environment. Because of their high ability to bioconcentration trace metals and organic compounds, bivalves have been widely applied for monitoring the status of and temporal changes in the trance contaminants, i.e., the Mussel Watch Project of the USA. While their response to chemical change in their surroundings may be detectable within a matter of days, depending on the species and on the contaminants, therefore, bio-accumulation kinetics of pollutants in marine life body is important. In the present paper, three kinds of marine bivalves (wild Saccostrea cucullata, aquacultured Perna viridis and Pinctada martens), collected from Daya Bay, the South China Sea, where the first nuclear power station of China has been running from 1994, were chose to investigate the bio-accumulation of radioruthenium from natural seawater of Daya Bay (pH 8.20, 35 per thousand salinity) under laboratory conditions. Before experiment, the individual bivalve was acclimated for 2 days and then thoroughly washed with seawater to remove surface-adsorbed or/and surface-attached substances which could be an important source of the uptake of particle-reactive radioruthenium. After each interval of experiment, 5 individuals of each kinds of bivalves were sampled, executed, and dissected. The soft tissue was dried, ashed and nitrated and the shell was dried and powdered. The correspondence activity was determined by a γ-ray spectrometer (EG and G ORTEC ADCAM-2000, P-type, relative efficiency 35%). Bio-concentration factor is defined by BCF (mL/g)=103Ru Activity in tissue of bivalves (Bq/g-fresh)/103Ru Activity in seawater (Bq/ml). The plots of BCF of ruthenium in the tissues of bivalves are shown as a function of time The results clearly show that BCF increase sharply with time from beginning to 6 days and approach a

  9. Preparation of Inkjet-Printed NiO Films for Ba(Ti,Zr)O3-Based Ceramics and Application to Multilayer Ceramics with Ni Electrodes

    Sakai, Yuichi; Futakuchi, Tomoaki; Adachi, Masatoshi

    2008-09-01

    The possibility of fabricating a lead-free multilayer ceramic (MLC) actuator with Ni inner electrodes prepared by inkjet printing has been investigated. Inkjet ink containing NiO powder was prepared. NiO films were prepared on [(BaO)1.00(CaO)0.01](Ti0.95Zr0.05)O2 green sheets by the inkjet method and co-fired in a reducing atmosphere. After co-firing, the NiO films were reduced to metal Ni films, which acted as electrodes. The remanent polarization Pr and coercive field Ec of the ceramics were 2.5 µC/cm2 and 3.0 kV/cm, respectively. The Curie temperature, orthorhombic-tetragonal transition temperature, and rhombohedral-orthorhombic transition temperature were 110, 45, and 0 °C, respectively. The [(BaO)1.00(CaO)0.01](Ti0.95Zr0.05)O2 MLCs with Ni inner electrodes were prepared using NiO ink. The diffusion and reaction of Ni to ceramic layers were not observed. The displacement of the MLCs with seven active layers was approximately 0.17 µm when the electric field was 20 kV/cm. It is expected that inkjet printing using NiO ink will be applicable to the fabrication of lead-free MLC actuators with Ni inner electrodes.

  10. Analysis and development of methods for the recovery of degraded tri-n-butyl phosphate (TBP)-30%V/V-dodecane

    Tri-n-butyl phosphate associated with an inert hydrocarbon, is the principal solvent used in reprocessing of nuclear irradiated fuel arising of pressurized water reactors, nowdays. The combined action of radiation and nitric acid cause severe damage to solvent, in reprocessing steps. Then, the recovery of solvent gets some importance, since it decreases the amount of the waste and improves the economy of the process. A comparative analysis of several methods of the recovery of this solvent was done, such as: alkaline washing, adsortion with resins, adsorption with aluminium oxide, adsorption by active carbon and adsorption by vermiculite. Some modifications of the analytical test of 95Zr and a mathematical definition of two new parameters were done: the degradation grade and the eficiency of recovering. Through this modified test of 95Zr, the residence time and the rate of degraded solvent: recuperator, were determined. After the laboratory tests had been performed, vermiculite, associated with active carbon, were employed in the treatment of 50 liters of tri-n-butyl phosphate (30%V/V)-dodecane, degraded by hydrolysis. Succeding analyses were made to check up the potentialities of these solids in the recovering of this solvent. (Author)

  11. Analysis and development of methods for the recovery of tri-n-butylphosphate (TBP)-30%v/v-degraded dodecane

    Tri-n-butyl phosphate associated with an inert hydrocarbon is the main solvent used in reprocessing of nuclear irradiated fuel arising of pressurized water reactors. The combined action of radiation and nitric acid cause severe damage to solvent, in reprocessing steps. The recovery of the solvent is, thus, of great importance, since it decreases the amount of the waste and improves the process economy. A comparative analysis of several methods of the recovery of this solvent was carried out, such as: alkaline washing, adsorption with resins, adsorption with aluminium oxide, adsorption by active carbon and adsorption by vermiculite. Some modifications of analytical 95Zr test and a mathematical definition of two new parameters (degradation grade and efficiency of recovery) were done. Through this modified 95Zr test, the residence time and the rate of degraded solvent: recuperator were determined. After laboratory tests, vermiculite associated with active carbon was employed for the treatment of 50 liters of tri-n-butyl phosphate (30% V/V)-dodecane, degraded by hydrolysis. Other analyses were performed to check the potentialities of these solids for this solvent recovery. (Author)

  12. Volatility of ruthenium-106, technetium-99, and iodine-129, and the evolution of nitrogen oxide compounds during the calcination of high-level, radioactive nitric acid waste

    The nitrate anion is the predominant constituent in all high-level nuclear wastes. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases (N2 or N2O), which can be scrubbed free of 106Ru, 129I, and 99Tc radioactivities prior to elimination from the plant by passing through HEPA filters. Treatment of a high-level authentic radioactive waste with two moles of formic acid per mole of nitrate anion leads to a low RuO4 volatility of about 0.1%, which can be reduced to an even lower level of 0.007% on adding a 15% excess of formic acid. Without pretreatment of the nitrate waste with formic acid, a high RuO4 volatility of approx. 35% is observed on calcining a 4.0 N HNO3 solution in quartz equipment at 3500C. The RuO4 volatility falls to approx. 1.0% on decreasing the initial HNO3 concentration to 1.0 N or lower. It is postulated that thermal denitration of a highly nitrated ruthenium complex leads to the formation of volatile RuO4, while decarboxylation of a ruthenium-formate complex leads to the formation of nonvolatile RuO2. Wet scrubbing with water is used to remove RuO4 from the off-gas stream. In all glass equipment, small amounts of particulate RuO2 are formed in the gas phase by decomposition of RuO4. The 99Tc volatility was found to vary from 0.2 to 1.4% on calcining HNO3 and HCOOH (formic acid) solutions over the temperature range of 250 to 6000C. These unexpectedly low volatilities of 99Tc are correlated to the high thermal stability limits of various metal pertechnetates and technetates. Iodine volatilities were high, varying from a low of 30% at 3500C to a high of 97% at 6500C. It is concluded that with a proper selection of pretreatment and operating conditions the 106Ru and 99Tc activities can be retained in the calcined solid with recycle of the wet scrubbing solution

  13. Exposure of the population of countries within the European Union to radioactivity in the mediterranean sea: Project MARINA-MED

    The discharges examined in the Project MARINA-MED were from nuclear installations in France, Italy and Spain, over the period 1980-1991; those from the Marcoule fuel reprocessing plant were found to contribute most of the collective dose arising from radioactivity in the Mediterranean Sea. Over 90% of beta-gamma discharges were of tritium and the remainder comprised some 80% 106Ru + 106Rh, 7% 90Sr + 90Y, 6% 137Cs, and 1% each of 134Cs, 58Co and 54Mn. The alpha emitters were dominated by 241Am, 239+240Pu and natural uranium, each representing 30% of the total. The contribution from Chernobyl had largely disappeared by 1990, except in the northern Aegean Sea, where continuing input came from the Black Sea. The corresponding exposure of the critical group in the Mediterranean Sea area from 137Cs (for an annual consumption of 73 kg fish and 35 kg shellfish) was estimated to be 7.5 μSv in 1990; the exposure of the critical group in the Black Sea area would have been about 40 μSv. These values are much lower than that estimated for 210Po, which corresponds to about 0.5 mSv. Data were obtained on catches and consumption rates of fish, crustaceans and molluscs for the relevant countries within the European Union, including import and export. Using data from the Food and Agriculture Organization of the United Nations, it was possible to extend the data to other countries bordering the Mediterranean Sea. Peak annual collective doses were estimated at less than 0.2 man·Sv for the period considered. For discharges over the entire period (1980-1991) the total collective dose commitment, truncated at 500 years, was estimated to be about 2 man·Sv. The most significant radionuclides, especially in the first few years, were found to be 106Ru (0.76 man·Sv) and 241Am (0.25 man·Sv). (author). 7 refs, 3 tabs

  14. Transmutation of 239Pu with spallation neutrons

    Incineration studies of plutonium were carried out at the Synchrophasotron of the Joint Institute for Nuclear Research (JINR), Dubna, using proton beams with energies of 0.53 GeV and 1.0 GeV. Solid lead targets (8 cm in diameter and 20 cm long) were surrounded with 6 cm thick paraffin as neutron moderator and then irradiated. The transmutation of 239Pu and the associated production of fission products 91Sr, 92Sr, 97Zr, 99Mo, 103Ru, 105Ru, 129Sb, 132Te, 133I, 135I and 143Ce were studied in the present work. The plutonium samples (each 449 mg) were placed on the outer surface of moderator. For 1.0 GeV proton beam, the fission rate of 239Pu is 0.0032 atoms per proton in one gram plutonium samples, for 0.53 GeV proton this value is 0.0022. The experimental uncertainty is about 15%. The experiments are compared to two theoretical model calculations with moderate success, using the Dubna Cascade Model (CEM) and the LAHET code. The practical incineration rate of 239Pu is very high. For example: if one uses 10 mA, 1 GeV proton beams under the same (fictive) experimental conditions, the incineration rate of 239Pu via fission is 3 mg out of the 449 mg sample per day. For 0.53 GeV protons the corresponding rate is 2 mg per day. (author)

  15. Monitoring on influence of Soviet chernobyl accident on environment of some regions of China

    This paper reports the monitoring results of some environmental samples from Gansu provinces and Qinshan aera of Zhejiang Province and the cities of Beijing, Shenyang and Baotou after the Soviet Chernobyl reactor accident. The samples collected included air, fallout, rain water, reservoir water, plants and soil and the wipping samples of international and domestic airlines were also measured. Analyese were made by using low background Ge(Li) γ spectrometer with anti-coincident shield and by radiochemical methods for 89Sr, 90Sr and Pu contents in some samples. The results indicate that the radioactive cloud released from the Chernobyl accident arrived to Beijing area on May 2, 1986. Generally speaking, the concentration of radioactive cloud in north China was greater than that in south China. Fission products were found in wipping samples taken from airplanes flying over Europe and Asia. The radioactivity level of the samples taken from European air-line was considerably higher than that from Asian airline. The main fission products found in different samples were as follows: 131I, 137Cs, 134Cs, 103Ru and 132Te, 132I. The ratio of 137Cs to 134Cs was about 2. The partial effective dose equivalent commitment of preliminary estimation to the public in Beijing area from the accident was 11.3 μSv. The contribution of the external exposure was 7.9 μSv. The contribution of the internal exposure was 3.4 μSv

  16. Production and radiochemical separation of rhodium-105 for radiopharmaceutical applications

    Production of no carrier added 105Rh by thermal neutron irradiation of natural Ru target and its radiochemical separation to obtain radionuclidically and radiochemically pure species is described. Irradiation planning, post irradiation chemical separation of 105Rh from ruthenium and iridium radionuclidic contaminants are discussed in detail. 100 mg of natural Ru was irradiated in a thermal neutron flux of 3 x 1013 neutrons/cm2/s for 7 days. Irradiated target was dissolved in a mixture of KIO4 and KOH in presence of water. Solvent extraction with CCl4 was used for removal of 97Ru and 103Ru, whereas solvent extraction with TBP was used for the removal of traces of 192Ir impurity. Different radiochemical separation techniques were tested for the recovery of 105Rh and its final purification from the accompanying radionuclidic impurities. The salts associated with the product were subsequently removed by fractional precipitation and by cation exchange chromatography. The radionuclidic purity of the 105Rh separated was estimated by gamma ray spectrometry; no radionuclidic contamination was observed. The radiochemical purity of the 105RhCl3 was ascertained by paper chromatography and paper electrophoresis. (orig.)

  17. Measurement of transit time of digesta through sections of gastrointestinal tract of pigs fed with diets containing various sources of dietary fibre (non-starch polysaccharides)

    The effects of various sources of dietary fibre (defined as non-starch polysaccharides (NSP)) on the transit time (TT) on digesta through sections of digestive tract were measured in pigs of 30-85 kg. The pigs were fitted with simple cannulas in the terminal ileum, caecum and mid-colon. Diets in experiments 1-3 were based on barley, wheat, soya bean meal and fish meal with NSP added in the form of wood cellulose (experiment 1), guar gum (experiment 2), wheat bran, pectin (experiment 3). Lactulose was also included in experiment 3 because of its NSP-like effects. Diets in experiments 4 and 5 were based on starch and casein and contained Phaseoluos vulgaris or Pisum sativum (experiment 4) and sugar beet pulp or wheat bran (experiment 5). Transit time (TT) was measured using 103Ru phenanthroline to label solids and 51Cr complexed to EDTA for liquids. Samples were taken every 3 h after marker administration for 51 h from all cannulas and the faecal output was collected every 3 h. The values obtained were very variable. The range of TT (h) defined as first arrival of markers and peak marker level was 3-12.2 and 3-12.2 to the ileum, 3-22.3 and 4.5-22.3 to the caecum, 4.5-50.3 and 16.5-48.8 to the colon and 24-<51 and 30-<51 to the rectum respectively. (author)

  18. Investigation of the α-particle induced nuclear reactions on natural molybdenum

    Complete text of publication follows. Cross-sections of alpha particle induced nuclear reactions on natural molybdenum have been studied in the frame of a systematic investigation of charged particle induced nuclear reactions on metals for different applications. The excitation functions of 93mTc, 93gTc(m+), 94mTc, 94gTc, 95mTc, 95gTc, 96gTc(m+), 99mTc, 93mMo, 99Mo(cum), 90Nb(m+), 94Ru, 95Ru, 97Ru, 103Ru and 88Zr were measured up to 40 MeV alpha energy by using a stacked foil technique and activation method. The main goals of this work were to get experimental data for accelerator technology, for monitoring of alpha beam, for thin layer activation technique and for testing nuclear reaction theories. The experimental data were compared with critically analyzed published data and with the results of model calculations, obtained by using the ALICE-IPPE, EMPIRE and TALYS codes (TENDL-2011). Yield versus energy curves are calculated from the measured data (Figs. 1-4) for the radioisotopes having special importance in one of the application fields.

  19. Investigation of the α-particle induced nuclear reactions on natural molybdenum

    Highlights: ► Excitation function measurement of α-particle induced reactions on natural molybdenum up to 40 MeV. ► Model code calculations with EMPIRE-II, EMPIRE3.1, ALICE and TALYS. ► Integral production yield calculation. ► Comparison with deuteron and proton production. ► Monitor reactions for α-irradiations. - Abstract: Cross-sections of alpha particle induced nuclear reactions on natural molybdenum have been studied in the frame of a systematic investigation of charged particle induced nuclear reactions on metals for different applications. The excitation functions of 93mTc, 93gTc(m+), 94mTc, 94gTc, 95mTc, 95gTc, 96gTc(m+), 99mTc, 93mMo, 99Mo(cum), 90Nb(m+), 94Ru, 95Ru,97Ru, 103Ru and 88Zr were measured up to 40 MeV alpha energy by using a stacked foil technique and activation method. The main goals of this work were to get experimental data for accelerator technology, for monitoring of alpha beam, for thin layer activation technique and for testing nuclear reaction theories. The experimental data were compared with critically analyzed published data and with the results of model calculations, obtained by using the ALICE-IPPE, EMPIRE and TALYS codes (TENDL-2011).

  20. Determination of Technetium-99 in Environmental Samples by Solvent Extraction at Controlled Valence

    Chen, Q.J.; Aarkrog, A.; Dahlgaard, H.;

    1989-01-01

    Distribution coefficients of technetium and ruthenium are determined under different conditions with CCl4, cyclohexanone, and 5% tri-isooctylamine (TIOA)/xylene. A method for analyzing 99Tc in environmental samples has been developed by solvent extraction in which the valences of technetium and...... subsequently separated by solvent extraction with cyclohexanone and 5% TIOA/xylene. The decontamination of the procedure is 1.35 .cntdot. 105 for 103Ru and 1.66 .cntdot. 105 for 110mAg. The chemical yield of technetium-99 is 55%....... ruthenium are controlled with H2O2 and NaClO. Technetium and ruthenium which are oxidized to TcO4- and RuO4- by NaClO are separated by extraction with CCl4 at pH 4. The RuO4- is reduced to low valence and technetium is kept in the TcO4- state with H2O2. Technetium, ruthenium, and other nuclides are...

  1. Radionuclide deposition and migration within the Gideaa and Finnsjoen study sites, Sweden: A study of the fallout after the Chernobyl accident

    Radionuclides originating from the Chernobyl accident in April 1986 were deposited over large areas of Sweden. The distribution and migration of the radionuclides during the first months after deposition were measured in a comprehensive survey within two study sites, Gideaa in Aangermanland county and Finnsjoen in Uppland county. The sites are previously investigated in the SKB site characterization programme and well defined regarding geology and hydrology. Radionuclides analysed are: Mn-54, Co-60, Sr-90, Zr-95, Nb-95, Mo-99, Ru-103, Ru-106, Ag-110m, Sb-125, I-131, Cs-134, Cs-136, Cs-137, Ba-140, La-140, Ce-141 and Ce-144. The CS-137 surface activity gave a range of 30-100 kBq/m2 in Gideaa and 20-40 kBq/m2 in Finnsjoen. Radionuclide migration is observed in soil profiles, groundwater and rock fissures. An active transport by surface water is also evident from sediment samples. Radionuclides have been absorbed in different types of vegetation. (orig./DG)

  2. Analytical procedure for 99Tc in forest soil by ICP-MS

    It is very important for 99Tc measurement by ICP-MS to remove 99Ru. So we examined the HNO3 concentration dependence with absorption about Tc and Ru to TEVA Resin which is used for Tc purification. The sample solution was acidified in 0.01 M HNO3 and passed through a column of 0.8 ml TEVA Resin. Ru and other matrix materials was removed by 1 M HNO3 and Tc was released by 10 M HNO3. The behavior of Ru through the soil sample preparation was examined using 103Ru tracer. Most of Ru was coprecipitated with Fe(OH)3 and removed. A small amount of Ru remaining in solutions was completely removed by purification steps using TEVA Resin. The decontamination factor of Ru was higher than 104 - 105, so it was confirmed that this procedure was very effective for the removal of Ru. The concentration of 99Tc in IAEA-375 reference materials was determined with this procedure as 0.19 ± 0.02 Bq/kg dry. The recovery of Tc was higher than conventional method. (author)

  3. Surrogate formulations for thermal treatment of low-level mixed waste. Part 1: Radiological surrogates

    Stockdale, J.A.D.; Bostick, W.D.; Hoffmann, D.P. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Lee, H.T. [Oak Ridge Associated Universities, TN (United States)

    1994-01-01

    The evaluation and comparison of proposed thermal treatment systems for mixed wastes can be expedited by tests in which the radioactive components of the wastes are replaced by surrogate materials chosen to mimic, as far as is possible, the chemical and physical properties of the radioactive materials of concern. In this work, sponsored by the Mixed Waste Integrated Project of the US Department of Energy, the authors have examined reported experience with such surrogates and suggest a simplified standard list of materials for use in tests of thermal treatment systems. The chief radioactive nuclides of concern in the treatment of mixed wastes are {sup 239}Pu, {sup 238}U, {sup 235}U, {sup 137}Cs, {sup 103}Ru, {sup 99}Tc, and {sup 90}Sr. These nuclides are largely by-products of uranium enrichment, reactor fuel reprocessing, and weapons program activities. Cs, Ru, and Sr all have stable isotopes that can be used as perfect surrogates for the radioactive forms. Technetium exists only in radioactive form, as do plutonium and uranium. If one wishes to preclude radioactive contamination of the thermal treatment system under trial burn, surrogate elements must be chosen for these three. For technetium, the authors suggest the use of natural ruthenium, and for both plutonium and uranium, they recommend cerium. The seven radionuclides listed can therefore be simulated by a surrogate package containing stable isotopes of ruthenium, strontium, cesium, and cerium.

  4. In vivo multitracer analysis technique. Screening of radioactive probes for noninvasive measurement of physiological functions in experimental animals

    A novel screening experiment, to find radioactive probes for non-invasive measurements of physiological functions in experimental animals, was tested using the in vivo multitracer analysis technique. The details of the efficiency of the detector settings used in the in vivo multitracer analysis technique were examined by both computer simulations and practical measurements. Multiple radioactive isotopes, i.e. multitracer, were prepared by irradiating a silver foil target with a heavy ion beam at the RIKEN ring cyclotron. After chemical separation of the silver target, the multitracer was finally dissolved in isotonic citrate buffer. The multitracer solution was intravenously injected into rats. Using a γ-ray detector equipped with a well-defined slit, the collimated γ-rays from the upper abdomen of living rats were measured. After correction of detection efficiencies, it was possible to compare the distribution of radioactive elements between two groups of rats different in body weight. The in vivo measurement showed that the tissue substantial volume of the selenium-deficient (SeD) rat liver increased compared to normal rats. The possibility of a functional estimation of tissue/blood volume for living rats was proposed based on the characteristic in vivo distribution of 74As, 83Rb and 103Ru. (author)

  5. Korean experimental studies on the radionuclide transfer in crop plants

    In Korea, data on the radionuclide transfer in crop plants have been produced almost exclusively at the Korea Atomic Energy Research Institute (KAERI), where experimental studies have been carried out for last about 20 years. These works are briefly outlined in this paper which shows results with emphasis on rice data. Soil-to-plant transfer factors of radionuclides including radiocesium and radiostrontium were measured through greenhouse experiments for various crop species. Not only conventional transfer factors but also those based on the activity applied to unit area of the soil surface were investigated. Field studies on the transfer of fallout 137Cs were carried out for rice and Chinese cabbage. As for parameters in relation to direct plant contamination, interception factors and translocation factors were obtained through greenhouse experiments. Plants were sprayed with radioactive solutions containing 54Mn, 57Co, 85Sr, 103Ru and 134Cs at different growth stages. Experiments on the plant exposure to airborne HTO and I2 vapor were also carried out. The transfer parameters generally showed great variations with soils, crops, radionuclides and isotope application times. Most experiments were designed for acute releases of radioactivity but some results are applicable to steady-state conditions, too. Many of the produced data would be of use also in other countries including Japan. (author)

  6. A radioecological survey of Northern and Middle Adriatic Sea before and after the Chernobyl event (1979-1990)

    Investigations to determine the presence and distribution of some anthropogenic pre- and post-Chernobyl radionuclides and to evaluate their behaviour and transport within different environmental components were conducted between 1979 and 1990 in Northern and Middle Adriatic Sea. Although the Chernobyl accident introduced many radionuclides into this environment, most of them had very short half lives and only Ru-103, Ru-106, Ag-110m, Cs-134 and Cs-137 were fully investigated on a long term basis due to their long persistency and abundance. In particular, neutron activation products, such as Ag-110m and Cs-134, not present in the environment before the event, were fully detected for the first time in the ecosystem shortly after the contamination. The results highlight the distribution of the considered gamma emitting radioisotopes in samples of seawater, sediments, mixed plankton, icthyofauna and benthic macrofauna, and allow a complete mapping and assessment of the Adriatic radiocontamination after 10 years of intense and extended monitoring of the ecosystem. (Author)

  7. Radioactivity in the Baltic sea following the Chernobyl accident

    The brown alga Fucus vesiculosus L. has been used as a bioindicator for the investigation of the impact of the Chernobyl accident with respect to the spatial and temporal distribution of radionuclides in the Baltic sea. The investigations were performed in July 1986, about two months after the accident, and in August-September 1987. In July 1986 the gamma-emitting radionuclides Cs-134, Cs-137, Ru-103, Ru-106 and Ag-110m were detected in F. vesiculosus along the Swedish east, south and southwest coasts. The activity concentrations of Cs-137 varied from 600 Bq/kg dw at the northern most locality (Simpnaes) to 20-25 Bq/kg dw at the south east coast. In August-September 1987 the activity concentrations of radiocesium had increased with a factor 2-3 at most localities off the Swedish east coast, compared with the results from 1986. Regarding transuranics and Tc-99 the impact was small and we did not observe any increase of these radionuclides in the algae. The later effects of the radionuclide contamination in the Baltic Sea, primarily cesium, from Chernobyl were studied at one locality on the Swedish south coast from April 1987 to November 1988. A pronounced increase in the activity concentrations was observed during 1988 indicating an outflow of water, containing relatively higher levels of Chernobyl derived radionuclides, from the Baltic Sea. (au)

  8. Development of pulmonary vascular response to oxygen

    The ability of the pulmonary circulation of the fetal lamb to respond to a rise in oxygen tension was studied from 94 to 146 days of gestation. The unanesthetized ewe breathed room air at normal atmospheric pressure, followed by 100% oxygen at three atmospheres absolute pressure in a hyperbaric chamber. In eleven near-term lambs, fetal arterial oxygen tension (PaO2) increased from 25 to 55 Torr, which increased the proportion or right ventricular output distributed to the fetal lungs from 8 to 59%. In five very immature lambs fetal PaO2 increased from 27 to 174 Torr, but the proportion of right ventricular output distributed to the lung did not change. In five of the near-term lambs, pulmonary blood flow was measured. For each measurement of the distribution of blood flow, approximately 8 x 105 spheres of 15-μm diameter, labeled with either 153Gd, 113Sn, 103Ru, 95Nb, or 46Sc were injected. It increased from 34 to 298 ml · kg fetal wt-1 · min-1, an 8.8-fold increase. The authors conclude that the pulmonary circulation of the fetal lamb does not respond to an increase in oxygen tension before 101 days of gestation; however, near term an increase in oxygen tension alone can induce the entire increase in pulmonary blood flow that normally occurs after the onset of breathing at birth

  9. Surrogate formulations for thermal treatment of low-level mixed waste

    The evaluation and comparison of proposed thermal treatment systems for mixed wastes can be expedited by tests in which the radioactive components of the wastes are replaced by surrogate materials chosen to mimic, as far as is possible, the chemical and physical properties of the radioactive materials of concern. In this work, sponsored by the Mixed Waste Integrated Project of the US Department of Energy, the authors have examined reported experience with such surrogates and suggest a simplified standard list of materials for use in tests of thermal treatment systems. The chief radioactive nuclides of concern in the treatment of mixed wastes are 239Pu, 238U, 235U, 137Cs, 103Ru, 99Tc, and 90Sr. These nuclides are largely by-products of uranium enrichment, reactor fuel reprocessing, and weapons program activities. Cs, Ru, and Sr all have stable isotopes that can be used as perfect surrogates for the radioactive forms. Technetium exists only in radioactive form, as do plutonium and uranium. If one wishes to preclude radioactive contamination of the thermal treatment system under trial burn, surrogate elements must be chosen for these three. For technetium, the authors suggest the use of natural ruthenium, and for both plutonium and uranium, they recommend cerium. The seven radionuclides listed can therefore be simulated by a surrogate package containing stable isotopes of ruthenium, strontium, cesium, and cerium

  10. Transfer of radionuclides from maternal food to the fetus and nursing infants of minipigs

    Transfer of 110mAg, 58Co, 59Fe, 141Ce, 103Ru, 88Y, 85Sr, 51Mn, 134Cs, 152Eu, 95mTc, 75Se, 65Zn and 133Gd was investigated in utero and during lactation in minipigs given the radioactive material added to food from day 50 of pregnancy until the end of lactation. The paper presents selected results on Ag, Co, Fe, Sr, Mn, Cs, Ru and Y and Tc. Transfer was highest for Cs and, in haemopoietic tissues, for Fe. Lower transfer was found for Ag, Fe, Mn with some preference for certain tissues (Ag in brain and liver, Mn in pancreas). Sr accumulated almost exclusively in bone and Tc in thyroid with higher concentrations in fetal and infant tissues than in maternal tissues. Lanthanides, Ru and Y were all close to detection limits or below in most maternal or fetal or infant tissues and could be found in bone and, less consistently, in kidney and liver. (author)

  11. Recent changes in liquid radioactive waste discharges to the Irish Sea from Sellafield. Pt.1. Inputs and uptake by coastal biota

    A revised authorization to discharge liquid radioactive waste from Sellafield was granted in January 1994. The revision took account of the need to review authorizations regularly, new radiological protection criteria, and new plant developments at Sellafield. These developments included the Thermal Oxide Reprocessing Plant (THORP) and the Enhanced Actinide Removal Plant (EARP). The revised authorization included reduced discharge limits for radionuclides of greatest radiological significance, such as 106Ru, 137Cs and the transuranics. To allow operation of the new plants, increases in discharge limits were allowed for some nuclides of low radiological significance, such as 3H, 14C, 99Tc and 129I. Changes in the discharges of radionuclides with the operation of the new developments are described, together with selected results from a comprehensive programme of aquatic environmental monitoring. The trends in results of monitoring over the period 1990-95 are related to the changes in discharge patterns since the new developments commenced operation, taking account of earlier discharges from Sellafield. (author)

  12. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO4(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO4(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO4(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO4(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for 106Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  13. Extraction of certain radionuclides from aqueous schungite solutions

    The sorption of 90Sr, 106Ru, 137Cs, and 238Pu from aqueous solutions over a wide pH range was studied. Swelled schungite chips (Nigozero, Kondopozhsk region) (1) and schungite (Onezhsk lake) (2) were tested as sorbents. The minerals were used both untreated and after oxidation (HNO3, 1:1, contact time 1 day). The oxidation, judging from the literature, facilitates the formation of carboxylic and phenolic functional groups on the surface of the carbon-containing sorbents. The presence of such groups is responsible for the high selectivity of the sorbents for multicharged cations. Futhermore, the hydrophobicity of the schungites enormously decreases after the oxidation. The studied sorbents had an average particle size of 0.08-0.1 mm. The schungite was contacted with the solution under static conditions with periodic stirring in order to establish equilibrium. The concentration of the radionuclides was 2-4 MBq/liter. The solution volumes were 10 ml. The sorbent content was 0.01 g. The required pH was set by adding HCl or NaOH

  14. Correction factors for the ISO rod phantom, a cylinder phantom, and the ICRU sphere for reference beta radiation fields of the BSS 2

    The International Organization for Standardization (ISO) requires in its standard ISO 6980 that beta reference radiation fields for radiation protection be calibrated in terms of absorbed dose to tissue at a depth of 0.07 mm in a slab phantom (30 cm x 30 cm x 15 cm). However, many beta dosemeters are ring dosemeters and are, therefore, irradiated on a rod phantom (1.9 cm in diameter and 30 cm long), or they are eye dosemeters possibly irradiated on a cylinder phantom (20 cm in diameter and 20 cm high), or area dosemeters irradiated free in air with the conventional quantity value (true value) being defined in a sphere (30 cm in diameter, made of ICRU tissue (International Commission on Radiation Units and Measurements)). Therefore, the correction factors for the conventional quantity value in the rod, the cylinder, and the sphere instead of the slab (all made of ICRU tissue) were calculated for the radiation fields of 147Pm, 85Kr, 90Sr/90Y, and, 106Ru/106Rh sources of the beta secondary standard BSS 2 developed at PTB. All correction factors were calculated for 0° up to 75° (in steps of 15°) radiation incidence. The results are ready for implementation in ISO 6980-3 and have recently been (partly) implemented in the software of the BSS 2

  15. Measurement of radionuclides in bricks made from refuse ash of sewage sludge

    Since people spend about 80% of their time in buildings, in order to estimate the radiation exposure dose of nation, it is necessary to grasp indoor dose level. It has been known that as for indoor radiation exposure dose, it is different according to the types of buildings, and generally it is low in wooden houses, and high in concrete and brick buildings. This difference is due to the shielding of environmental radiation by building material and the radiation emitted from building materials themselves. In this study, the radionuclides in the bricks made of sewage slude-incinerated ash and red bricks for the reference were investigated. The radiation emitted from the bricks was measured with a thermoluminescence dosimeter (TLD). The sampling and pretreatment, the measuring method, the measurement of the radiation dose on the bricks, and the results are reported. As for the natural radionuclides in the bricks, 7Be was detected in most sludge bricks, and the concentration varied largely according to the data of sampling. It was not detected in red bricks. 40K, 214Bi and 228Ac were detected in all the sludge and red bricks. Artificial radionuclides such as 54Mn, 60Co, 106Ru, 131I, 134Cs and 144Ce were not detected in all the sludge and red bricks. The radiation dose on the sludge and red bricks was 20.5 - 21.7 mR, and constant regardless of the data of sampling. (K.I.)

  16. The transfer of radionuclides from saltmarsh vegetation to sheep tissues and milk

    Radionuclides released into the Irish Sea by the Sellafield reprocessing plant are deposited onto tide-washed pastures along the western coast of the United Kingdom. Many of these pastures are grazed by sheep or cattle. This paper describes a controlled feeding study, in which saltmarsh vegetation harvested from close to the Sellafield plant, was fed to lambs and adult female sheep for a period of 8 weeks. Activity concentrations of 60Co, 95Nb, 106Ru, 134Cs, 137Cs, 238Pu, 239,240Pu and 241Am were determined in edible tissues and transfer parameters estimated. The activity concentrations of some of the radionuclides will not have been in equilibrium with those in the diet. Nevertheless, the study was reasonably realistic in terms of agricultural management as the period of the study was similar to that for which lambs graze on the saltmarshes. A field study to determine the activity concentrations of 137Cs and 239,240Pu in the milk of ewes grazing a saltmarsh close to Sellafield is also described

  17. Radiation Protection in Brachytherapy. Report of the SEFM Task Group on Brachytherapy; Proteccion radiologica en Braquiterapia. Informe del grupo de trabajo de Braquiterapia de la SEFM

    Perez-Calatayud, J.; Corredoira Silva, E.; Crispin Contreras, V.; Eudaldo Puell, T.; Frutos Baraja, J. de; Pino Sorroche, F.; Pujades Claumarchirant, M. C.; Richart Sancho, J.

    2015-07-01

    This document presents the report of the Brachytherapy Task Group of the Spanish Society of Medical Physics. It is dedicated to the radiation protection aspects involved in brachytherapy. The aim of this work is to include the more relevant aspects related to radiation protection issues that appear in clinical practice, and for the current equipment in Spain. Basically this report focuses on the typical contents associated with high dose rate brachytherapy with {sup 1}92Ir and {sup 6}0Co sources, and permanent seed implants with {sup 1}25I, {sup 1}03Pd and {sup 1}31Cs, which are the most current and widespread modalities. Ophthalmic brachytherapy (COMS with {sup 1}25I, {sup 1}06Ru, {sup 9}0Sr) is also included due to its availability in a significant number of spanish hospitals. The purpose of this report is to assist to the medical physicist community in establishing a radiation protection program for brachytherapy procedures, trying to solve some ambiguities in the application of legal requirements and recommendations in clinical practice. (Author)

  18. Field and model investigations of external gamma dose rates along the Cumbrian coast, NW England.

    McDonald, P; Bryan, S E; Hunt, G J; Baldwin, M; Parker, T G

    2005-03-01

    A survey of the contribution to external dose from gamma rays originating from intertidal sediments in the vicinity of the British Nuclear Group Sellafield site showed that the major anthropogenic contributions were due to (137)Cs and (60)Co. At some sites, traces of other anthropogenic radionuclides were detected, namely (106)Ru, (125)Sb, and (154)Eu. The proportions of fine grained material (Drigg Barn Scar and Whitehaven Coal Sands sites, which had their own unique characteristics. The highest (60)Co activity concentrations in this study were detected at Drigg Barn Scar. These relatively high activity concentrations of (60)Co were due to the presence of (60)Co in mussels and barnacles, hence upsetting the fine sediment relationships used in previous dose calculations. Whitehaven Coal Sands was unusual in that it contained higher levels of radionuclides than would be expected in sandy sediment. The mineralogy of these sediments was the controlling factor on (137)Cs binding, rather than the proportion of fine grained material. By adjusting the effective fine grained sediment proportions for calculations involving (60)Co and (137)Cs at Drigg Barn Scar and Whitehaven Coal Sands respectively, the CUMBRIA77/DOSE77 model predictions could be improved upon significantly for these sites. This work highlights the influence of particle size and sediment composition on external dose rate calculations, as well as the potential for external dose contributions from biota. PMID:15798279

  19. Field and model investigations of external gamma dose rates along the Cumbrian coast, NW England

    A survey of the contribution to external dose from gamma rays originating from intertidal sediments in the vicinity of the British Nuclear Group Sellafield site showed that the major anthropogenic contributions were due to 137Cs and 60Co. At some sites, traces of other anthropogenic radionuclides were detected, namely 106Ru, 125Sb, and 154Eu. The proportions of fine grained material (60Co activity concentrations in this study were detected at Drigg Barn Scar. These relatively high activity concentrations of 60Co were due to the presence of 60Co in mussels and barnacles, hence upsetting the fine sediment relationships used in previous dose calculations. Whitehaven Coal Sands was unusual in that it contained higher levels of radionuclides than would be expected in sandy sediment. The mineralogy of these sediments was the controlling factor on 137Cs binding, rather than the proportion of fine grained material. By adjusting the effective fine grained sediment proportions for calculations involving 60Co and 137Cs at Drigg Barn Scar and Whitehaven Coal Sands respectively, the CUMBRIA77/DOSE77 model predictions could be improved upon significantly for these sites. This work highlights the influence of particle size and sediment composition on external dose rate calculations, as well as the potential for external dose contributions from biota

  20. Assessment of the potential for radionuclide migration from a nuclear-explosion cavity

    The source term for radionuclides in the region of the Cambric nuclear explosion has been determined. Drillback cores were obtained and analyzed, and water was pumped from several vertical zones and analyzed. Most of the radioactivity produced in the test was found to be retained in the fused debris with only low concentrations in the water which had been in contact with the debris for nearly ten years. Most of the radioactivity and the highest specific activities of all radionuclides were found to be in the region of the original explosion cavity. No activity was found 50 m below the cavity. Water from the region of highest radioactivity at the bottom of the cavity contained only tritium and 90Sr at levels higher than the recommended concentration guides for drinking water in uncontrolled areas. During nearly six years of pumping from a satellite well located 91 m from the Cambric cavity, only tritium 85Kr have been positively identified in water removed from this well, although there is some evidence for the possible migration of minute amounts of 106Ru. These results are consistent with laboratory studies which indicate that, in general, radionuclide sorption is sufficiently high to preclude the migration of such nuclides from the original cavity to the satellite well in the near future

  1. Development of a CMPO based extraction process for partitioning of minor actinides and demonstration with genuine fast reactor fuel solution (155 GWd/Te)

    Antony, M.P.; Kumaresan, R.; Suneesh, A.S. [Indira Gandhi Centre for Atomic Research, Kalpakkam (IN). Fuel Chemistry Div.] (and others)

    2011-07-01

    A method has been developed for partitioning of minor actinides from fast reactor (FR) fuel solution by a TRUEX solvent composed of 0.2 M n-octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (CMPO)-1.2 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), and subsequently demonstrated with genuine fast reactor dissolver solution (155 GWd/Te) using a novel 16-stage ejector mixer settler in hot cells. Cesium, plutonium and uranium present in the dissolver solution were removed, prior to minor actinide partitioning, by using ammonium molybdophosphate impregnated XAD-7 (AMP-XAD), methylated poly(4-vinylpyridine) (PVP-Me), and macroporous bifunctional phosphinic acid (MPBPA) resins respectively. Extraction of europium(III) and cerium(III) from simulated and real dissolver solution, and their stripping behavior from loaded organic phase was studied in batch method using various citric acid-nitric acid formulations. Based on these results, partitioning of minor actinides from fast reactor dissolver solution was demonstrated in hot cells. The extraction and stripping profiles of {sup 154}Eu, {sup 144}Ce, {sup 106}Ru and {sup 137}Cs, and mass balance of {sup 241}Am(III) achieved in the demonstration run have been reported in this paper. (orig.)

  2. Development of a CMPO based extraction process for partitioning of minor actinides and demonstration with genuine fast reactor fuel solution (155 GWd/Te)

    A method has been developed for partitioning of minor actinides from fast reactor (FR) fuel solution by a TRUEX solvent composed of 0.2 M n-octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (CMPO)-1.2 M tri-n-butylphosphate (TBP) in n-dodecane (n-DD), and subsequently demonstrated with genuine fast reactor dissolver solution (155 GWd/Te) using a novel 16-stage ejector mixer settler in hot cells. Cesium, plutonium and uranium present in the dissolver solution were removed, prior to minor actinide partitioning, by using ammonium molybdophosphate impregnated XAD-7 (AMP-XAD), methylated poly(4-vinylpyridine) (PVP-Me), and macroporous bifunctional phosphinic acid (MPBPA) resins respectively. Extraction of europium(III) and cerium(III) from simulated and real dissolver solution, and their stripping behavior from loaded organic phase was studied in batch method using various citric acid-nitric acid formulations. Based on these results, partitioning of minor actinides from fast reactor dissolver solution was demonstrated in hot cells. The extraction and stripping profiles of 154Eu, 144Ce, 106Ru and 137Cs, and mass balance of 241Am(III) achieved in the demonstration run have been reported in this paper. (orig.)

  3. Waste/Rock Interactions Technology Program. Status report on LWR spent-fuel leach tests

    Spent fuels with burnups of 9000, 28,000 and 54,500 MWd/MTU have been leach tested at 250C. Three leach-test procedures (Paige, IAEA and static) were used. IAEA and static tests were conducted in five different solutions: deionized water, sodium bicarbonate, sodium chloride, calcium chloride and Waste Isolation Pilot Plant B brine solutions. Elemental leach data are reported based on the release of 90Sr+90Y, 106Ru, 137Cs, 144Ce, 154Eu, 239+240Pu, 125Sb, 244Cm, 129I, 99Tc, and total uranium. This is the first report on 129I and 99Tc from spent fuel. Termination of the Paige test showed that the plateout (radionuclide adsorption) on the test apparatus had negligible effect on the leach rate of cesium and plutonium, but a major (up to a factor of 50 times) effect on the curium leach rate. Three-hundred additional days of leach testing by the IAEA procedure, from 467 to 769 d, showed a continuation of the leaching trends observed during the first 467 d. Results from the first two static leach test series, 2 and 8 d, gave the 129I and 99Tc release numbers

  4. Extrapolation of experimental data on late effects of low-dose radionuclides in man

    The situation of living of population on radionuclide contamination areas was simulated in the experimental study using white strainless rats of different ages. The significance of age for late stochastic effects of internal radionuclide contamination with low doses of 131I, 137Cs, 144Ce and 106Ru was studied. Some common regularities and differences in late effects formation depending on age were found. Results of the study showed that the number of tumors developed increased in groups of animals exposed at the youngest age. The younger animal at the moment of internal radionuclide contamination, the higher percentage of malignant tumors appeared. It was especially so for tumors of endocrine glands (pituitary, suprarenal,- and thyroid). Differences in late effects formation related to different type of radionuclide distribution within the body were estimated. On the base of extrapolation the conclusion was made that human organism being exposed at early postnatal or pubertal period could be the most radiosensitive (1.5-2.0 or sometimes even 3-5 times higher than adults). Data confirmed the opinion that children are the most critical part of population even in case of low dose radiation exposure. (author)

  5. Contribution to investigations on trace elements transport in the Channel: spatial distribution of industrial tracers in mytilus edulis and fucus serratus

    The distribution of artificial tracers - gamma emitters - has been studied in biological indicator species, mussels and fucus, along the french and english Channel shores in order to gain a better knowledge of trace elements transports in the Channel coastal areas. The main conclusions are supplied by 106Ru-Rh and 60Co. Extension of species labelling is larger eastwards than westwards, and the differences recorded between french and english shores show weak exchanges between south and north Channel; in the norman-breton gulf and in the Seine river bay, the distribution of radioactive tracers demonstrates complex current processes. The results are compared to the hydrodynamical studies carried out through models and follow-up of radioactive tracers in sea-water. Particular processes have been observed, corresponding to areas where the decay gradient from the source term is not respected (western Cotentin shore, western Seine Bay, Caux aerea). They are discussed in relation with fresh - sea water mixing, current and physico-chemical problems

  6. A study of the decontamination effect of commercial detergents on the skin

    Titanium dioxide paste is generally used as the radiological skin decontaminant in the radiation control area at Japan Nuclear Cycle Development Institute (JNC) Tokai works. It is a typical and proven skin decontaminant in the nuclear industry, but there is a disadvantage in that it has a short shelf life. Recently, many detergents applicable as skin decontaminants have become available in non-nuclear industries such as cosmetics and sanitation. They are easily acquired and have an advantage in that stimulus to the skin is mild, because these products have been developed for the human body. In this study, the decontamination factor of the commercial detergents for each nuclide was examined using imitation skin. Sheets of raw pig skins were contaminated with a nitric-acid solution containing 144Ce, 137Cs, 106Ru or 60Co, and then washed with various detergents such as a neutral detergent, cleansing cream and orange oil. The nuclide removal rates of some of the commercial detergents examined were nearly equal to that of titanium dioxide, thus proving that they show satisfactory decontamination performance as a skin decontaminant. (author)

  7. The present state of research on the vitrification of concentrated solutions of fission products (1962)

    The present report gives the actual point of studies on vitrification of concentrated solutions of fission products. An active cell, giving glasses in crucibles, permitted to study various glass compositions. The leaching rate from the glass raises 1 to 2 10-7 g of glass/cm2/day. Activity loss by volatility during vitrification remains weak and often below 0.1 per cent of total activity. Off gas cleaning is made easier by presence of filter which is compound of granules including iron oxide. After saturation the content of this filter can be melt. Moreover different processes are in experimentation for a more important production. Daily 72 liters of solution containing tracer activity are treated in a continuous calcination and vitrification plant. The loss in 106Ru is still important and a modification of installation has been necessary. A pot vitrification plant is in study. In order to reduce cost of processing the possibility to pour glass after melting is actuality in study. A production set of very active glass is also in project. (authors)

  8. Study of charge collection and noise in non-irradiated and irradiated silicon detectors

    Leroy, C; Dezillie, B; Glaser, M; Lemeilleur, F; Trigger, I

    1996-01-01

    The large collection and noise were studied in non-irradiated and irradiated silicon detectors as a function of temperature (T), shaping time (0) and fluence , up to about 1,2 x 10(14) protons per cm2 for minimum-ionizing electrons yielded by a 106 Ru source. The noise of irradiated detectors is found to be dominted for short shaping times (¾50ns) by a series noise compo- nent, while for longer shaping times („80ns) a parallel noise component (correlated with the reverse current) prevails. For non-irradiated detectors, where the reverse current is three orders of magnetude smaller compared with irradiated detectors, the series noises dominates over the whole range of shaping times investigated (20-150ns). A signal degradation is observed for irradiated detectors. However, the signal ca be distinguished from noise, even after a fluence of about 1.2 x10(14) protons per cm2, at a temperature of 6øC and with a shaping time tipical of rge LHC inter-bunch crossing time (20-30ns). The measurements of the signal a...

  9. Study of charge collection and noise in non-irradiated and irradiated silicon detectors

    Leroy, C. [Montreal Univ., PQ (Canada); Bates, S. [CERN, ECP, CH-1211 Geneva (Switzerland); Dezillie, B. [CERN, ECP, CH-1211 Geneva (Switzerland); Glaser, M. [CERN, ECP, CH-1211 Geneva (Switzerland); Lemeilleur, F. [CERN, ECP, CH-1211 Geneva (Switzerland); Trigger, I. [Montreal Univ., PQ (Canada)

    1997-04-01

    Charge collection and noise were studied in non-irradiated and irradiated silicon detectors as a function of temperature (T), shaping time ({theta}) and fluence ({Phi}), up to about 1.2 x 10{sup 14} protons cm{sup -2}, for minimum ionizing electrons yielded by a {sup 106}Ru source. The noise of irradiated detectors is found dominated for short shaping times ({theta}{<=}50 ns) by a series noise component while for longer shaping time ({theta}{>=}80 ns) a parallel noise component (correlated with the reverse current) prevails. For non-irradiated detectors, where the reverse current is three orders of magnitude smaller compared with irradiated detectors, the series noise dominates over the whole range of shaping times investigated (20-150 ns). A signal degradation is observed for irradiated detectors. However, the signal can be distinguished from noise, even after a fluence of about 1.2 x 10{sup 14} protons cm{sup -2}, at a temperature of 6 C and with a shaping time typical of LHC inter-bunch crossing time (20-30 ns). The measurement of the signal as a function of voltage shows that irradiated detectors depleted at 50% of the full depletion voltage can still provide a measurable signal-to-noise ratio. (orig.).

  10. Environment surveillance

    The Cogema group involved in the nuclear fuel cycle from the uranium extraction to the reprocessing of spent fuels and the recycling of re-usable energy materials, performs periodical controls of radioactivity levels in the environment close to its facilities. At Cogema-La Hague, 80000 analyses were performed on 25000 samples taken at 820 different sites in the vicinity of the facility in 1997. These controls concern different elements of the terrestrial and maritime environments (surface and ground waters, rain water, plants, beach sand, marine sediments, marine fauna and flora). After a brief presentation of natural and artificial radioactivity, this brochure presents the results of the measurements performed by Cogema-La Hague (alpha- and beta-activity, tritium, 7Be, 40K, 241Am, 137Cs, 125Sb, 129I, 90Sr, 106Ru, 60Co, 110mAg) in air, surface waters, grass, milk, coastal waters, algae, molluscs and fishes. A special chapter is devoted to the presentation of tritium (origin, metabolism, toxicity, concentration in effluents). (J.S.)

  11. Interaction of radionuclides with geomedia associated with the Waste Isolation Pilot Plant (WIPP) site in New Mexico

    A survey of the potential of geological media from the vicinity of the Waste Isolation Pilot Plant site in Southeastern New Mexico for retardation of radionuclide migration in an aqueous carrier was conducted. The survey included the measurement of sorption coefficients (Kd) for twelve radionuclides between three natural water simulants and ten samples from various geological strata. The nuclides included 137Cs, 85Sr, 131I, 99Tc, 125Sb, 144Ce, 152Eu, 153Gd, 106Ru, 243Am, 244Cm, and 238Pu. The compositions of the simulant solutions were those expected of water in contact with potash minerals or halite deposits in the area and in a typical groundwater found in the Delaware Basin. The geological samples were obtained from potential aquifers above and below the proposed repository horizons and from bedded salt deposits in the repository horizons. In brine solutions, Tc and I were not significantly adsorbed by any of the minerals and Cs and Sr showed minimal adsorption (Kd's 103 and Ru and Sb Kd's varied in the range of 25 to > 103. In the groundwater simulant, Tc and I showed the same behavior, but the Kd's of the other nuclides were generally higher. Some initial parametric studies involving pH, trace organic constituents in the simulant solutions, and radionuclide concentrations were carried out. Significant differences in the observed Kd's can result from varying one or more of these solution parameters

  12. Innovative separation method for advanced spent fuel reprocessing based on tertiary pyridine resin

    Radiochemical separation experiments have been performed in order to realize a novel reprocessing method based on chromatography techniques using a novel pyridine resin. The newly synthesized tertiary pyridine resin with two functions of ion exchanger and soft-donor was dedicated to the experiments, where highly irradiated mixed oxide fuel from the experimental fast reactor JOYO was used as a reference spent fuel. With a 3-step separation, pure Am and Cm were individually obtained as minor actinide products, and 106Ru group, lanthanides with 137Cs group and Pu group were fractionated, respectively. The decontamination factor of 137Cs and trivalent lanthanides (155Eu, 144Ce) against the Am product exceeded 3.9 x 104 and 1.0 x 105, respectively. The decontamination factor as the mutual separation of 243Cm was larger than 2.2 x 103 against the Am product. Moreover, the content of 137Cs, trivalent lanthanides and 243Cm in Am product did not exceed 2 ppm. The tertiary pyridine resin method is a candidate separation system for an 'advanced ORIENT process', where enhanced separation, transmutation and utilization of actinides, long-lived fission products and rare metal fission product would be oriented. (author)

  13. Chemical and physical considerations of the use of nuclear fuel spikants for deterrence

    Selle, J. E.

    1978-10-01

    One proposed method of inhibiting the diversion of nuclear fuel for clandestine purposes is to add to the fuel a highly gamma-active material of such intensity that remote handling equipment is necessary in all stages of handling and reprocessing. This is called spiking for deterrence. The present work sought to identify candidate spikants and identify potential materials problems that might occur as the result of incorporation of these spikants with the fuel. Potential spikants were identified and thermodynamic analysis was performed to determine their chemical and physical states. Phase relationships between spikants (and their decay products) and the fuel constituents were surveyed. According to criteria defined in this report, /sup 60/Co, /sup 106/Ru, and /sup 144/Ce appear to have the greatest potential as spikants. Cerium should be present as the oxide, soluble in the fuel, while cobalt and ruthenium should be present in the metallic state with very low solubility in the fuel. Experimental work on the distribution of fission products and their interactions with cladding was also surveyed to provide information on the distribution of spikants in the fuel and describe the probable effects of spikants on the fuel. Cobalt, ruthenium, and cerium should not present any problems due to reaction with stainless steel cladding.

  14. Electrochemical treatment of liquid wastes

    Hobbs, D.T. [Savannah River Technology Center, Aiken, SC (United States)

    1997-10-01

    Under this task, electrochemical treatment processes are being evaluated and developed for the destruction of organic compounds and nitrates/nitrites and the removal of other hazardous species from liquid wastes stored throughout the DOE complex. This technology targets the (1) destruction of nitrates, nitrites and organic compounds; (2) removal of radionuclides; and (3) removal of RCRA metals. The development program consists of five major tasks: (1) evaluation of electrochemical reactors for the destruction and removal of hazardous waste components, (2) development and validation of engineering process models, (3) radioactive laboratory-scale tests, (4) demonstration of the technology in an engineering-scale reactor, and (5) analysis and evaluation of test data. The development program team is comprised of individuals from national laboratories, academic institutions, and private industry. Possible benefits of this technology include: (1) improved radionuclide separation as a result of the removal of organic complexants, (2) reduction in the concentrations of hazardous and radioactive species in the waste (e.g., removal of nitrate, mercury, chromium, cadmium, {sup 99}Tc, and {sup 106}Ru), (3) reduction in the size of the off-gas handling equipment for the vitrification of low-level waste (LLW) by reducing the source of NO{sub x} emissions, (4) recovery of chemicals of value (e.g. sodium hydroxide), and (5) reduction in the volume of waste requiring disposal.

  15. Electrochemical treatment of liquid wastes

    Under this task, electrochemical treatment processes are being evaluated and developed for the destruction of organic compounds and nitrates/nitrites and the removal of other hazardous species from liquid wastes stored throughout the DOE complex. This technology targets the (1) destruction of nitrates, nitrites and organic compounds; (2) removal of radionuclides; and (3) removal of RCRA metals. The development program consists of five major tasks: (1) evaluation of electrochemical reactors for the destruction and removal of hazardous waste components, (2) development and validation of engineering process models, (3) radioactive laboratory-scale tests, (4) demonstration of the technology in an engineering-scale reactor, and (5) analysis and evaluation of test data. The development program team is comprised of individuals from national laboratories, academic institutions, and private industry. Possible benefits of this technology include: (1) improved radionuclide separation as a result of the removal of organic complexants, (2) reduction in the concentrations of hazardous and radioactive species in the waste (e.g., removal of nitrate, mercury, chromium, cadmium, 99Tc, and 106Ru), (3) reduction in the size of the off-gas handling equipment for the vitrification of low-level waste (LLW) by reducing the source of NOx emissions, (4) recovery of chemicals of value (e.g. sodium hydroxide), and (5) reduction in the volume of waste requiring disposal

  16. Neutron cross sections of 28 fission product nuclides adopted in JENDL-1

    This is the final report concerning the evaluated neutron cross sections of 28 fission product nuclides adopted in the first version of Japanese Evaluated Nuclear Data Library (JENDL-1). These 28 nuclides were selected as being most important for fast reactor calculations, and are 90Sr, 93Zr, 95Mo, 97Mo, 99Tc, 101Ru, 102Ru, 103Rh, 104Ru, 105Pd, 106Ru, 107Pd, 109Ag, 129I, 131Xe, 133Cs, 135Cs, 137Cs, 143Nd, 144Ce, 144Nd, 145Nd, 147Pm, 147Sm, 149Sm, 151Sm, 153Eu and 155Eu. The status of the experimental data was reviewed over the whole energy range. The present evaluation was performed on the basis of the measured data with the aid of theoretical calculations. The optical and statical models were used for evaluation of the smooth cross sections. An improved method was developed in treating the multilevel Breit-Wigner formula for the resonance region. Various physical parameters and the level schemes, adopted in the present work are discussed by comparing with those used in the other evaluations such as ENDF/B-IV, CEA, CNEN-2 and RCN-2. Furthermore, the evaluation method and results are described in detail for each nuclide. The evaluated total, capture and inelastic scattering cross sections are compared with the other evaluated data and some recent measured data. Some problems of the present work are pointed out and ways of their improvement are suggested. (author)

  17. Gamma spectrometrical analyses of annual deposition samples collected in Romania in 1987-1989

    The dynamics of artificial radionuclide concentrations in deposition samples in the years 1987-1989 is presented in this paper. High resolution gamma spectrometrical analyses have been performed on wet and dry deposition samples from the following stations of the National Environmental Radioactivity Surveillance Network: Cluj-Napoca, Satu-Mare, Constanta, Ceahlau-Tosca, lasi and Craiova. In order to determine small concentrations of low level artificial radionuclides (having an appropriate half life) yearly samples obtained by cumulating daily samples were measured. The following gamma emitting radionuclides were identified in these samples: 106Ru, 110mAg, 125Sb, 134Cs, 137Cs, 144Ce. It can be seen that the level of radionuclide activities in deposition samples at any time is dependent on the 'initial' deposition of May 1986. An environmental half life of 137Cs in deposition can be assessed. For a particular case the evolution during 1987-1989 of 137Cs and 7Be specific activities in monthly samples is also presented. (author)

  18. Beta Radiation Exposure of Personnel in Radiosynovectomy and at the Production of Eye Application

    Full text: Beta radionuclides are increasingly used in nuclear medicine therapy. In a study of exposure at working places with supposed enhanced radiation risk the followings are monitored: 1. Production of eye applicators (106Ru/106Rh) for therapy of intraocular tumours. During the processes radionuclides are used in sealed and unsealed form. 2. Radiosynovectomy (RSO). This therapy is often used in treatment of inflammatory joint disease. The radionuclides 169Er, 186Re and 90Y are applied in form of radioactive solutions. A complex of problems has to be solved here: Handling of high activities (several GBq per day), very small distances between source and skin, high risk of skin contamination, unsatisfactory dose measurement techniques. In our investigations high sensitive thin thermoluminescence dosimeters (LiF:Mg,Cu,P) were used to determine the skin dose of the hands especially fingertips during preparation and application of the radioactive substances at several institutions. During production of eye applicators we found hand doses essentially below the annual limit. But direct radiation at RSO caused more than 100 mSv at the fingertips per working day in some cases. Even higher doses (more than 50% of the annual limit for skin during a working day) were caused by contamination of the hands. By use of manipulators, wearing of appropriate protection gloves and by improvement of directions for work, the radiation exposure could be reduced dramatically. Consequences for individual beta dosimetry and with respect to license of use of beta radionuclides are discussed. (author)

  19. Estimated radiological doses to the maximumly exposed individual and downstream populations from releases of tritium, strontium-90, ruthenium-106, and cesium-137 from White Oak Dam

    Concentrations of tritium, 90Sr, 106Ru, and 137Cs in the Clinch River for 1978 were estimated by using the known 1978 releases of these nuclides from the White Oak Dam and diluting them by the integrated annual flow rate of the Clinch River. Estimates of 50-year dose commitment to a maximumly exposed individual were calculated for both aquatic and terestrial pathways of exposure. The maximumly exposed individual was assumed to reside at the mouth of White Oak Creek where it enters the Clinch River and obtain all foodstuffs and drinking water at that location. The estimated total-body dose from all pathways to the maximumly exposed individual as a result of 1978 releases was less than 1% of the dose expected from natural background. Using appropriate concentrations of to subject radionuclides diluted downstream, the doses to populations residing at Harriman, Kingston, Rockwood, Spring City, Soddy-Daisy, and Chattanooga were calculated for aquatic exposure pathways. The total-body dose estimated for aquatic pathways for the six cities was about 0.0002 times the expected dose from natural background. For the pathways considered in this report, the nuclide which contributed the largest fraction of dose was 90Sr. The largest dose delivered by 90Sr was to the bone of the subject individual or community

  20. Development of a method for analyzing traces of ruthenium in plant materials and determination of the transfer factors soil/plant for ruthenium compounds from reprocessing plants

    In an artificial humous and sandy soil spiked with 106Ru as RuO2 and RuCl3, pasture grass was grown under artificial illumination in our laboratory. The amounts of ruthenium taken up by the plants were determined by γ-spectrometry. For open-air investigations with pasture grass, wheat and potatoes inactive ruthenium(III) chloride and ruthenium nitrosylchloride were used. Ruthenium was determined by electrothermal atomic absorption spectrometry (ETAAS) after destroying the organic material and concentrating the solution. The concentration and chemical form of the ruthenium exert an unimportant influence on the transfer factor. For the pasture-grass, the stems of wheat and the weed of potatoes it amounts to 0.00005 to 0.0015, for the ear of wheat to about 0.00005. In peeled potatoes there was no ruthenium detectable, therefore the limit of detection leads to a transfer factor ≤ 0.00001. So it is evident that ruthenium is little available for the roots of the plants. In the event of an accident in a nuclear plant the uptake of radioactive ruthenium by roots has only negligible radioecological consequences. This applies even if 50 years of ruthenium enrichment in the soil are assumed. (orig./RB)

  1. Chernobyl radionuclides in a Black Sea sediment trap.

    Buesseler, K O; Livingston, H D; Honjo, S; Hay, B J; Manganini, S J; Degens, E; Ittekkot, V; Izdar, E; Konuk, T

    The Chernobyl nuclear power station accident released large quantities of vaporized radionuclides, and, to a lesser extent, mechanically released small (less than 1-10 micron) aerosol particles. The total release of radioactivity is estimated to be out of the order of 1-2 x 10(18) Bq (3-5 x 10(7) Ci) not allowing for releases of the xenon and krypton gases. The 137Cs releases of 3.8 x 10(16) Bq from Chernobyl can be compared to 1.3 x 10(18) Bq 137Cs released due to atmospheric nuclear weapons testing. Chernobyl-derived radionuclides can be used as transient tracers to study physical and biogeochemical processes. Initial measurements of fallout Chernobyl radionuclides from a time-series sediment trap at 1,071 m during June-September 1986 in the southern Black Sea are presented. The specific activities of 137Cs, 144Ce and 106Ru in the trap samples (0.5-2, 4-12 and 6-13 Bq g-1) are independent of the particle flux while their relative activities reflect their rates of scavenging in the order Ce greater than Ru greater than Cs. PMID:3670387

  2. Chemical isotopic analysis of fission products in PWR-MOX spent fuels and computational evaluation using JENDL, ENDF/B, JEF, and JEFF

    A chemical isotopic analysis of high-burn-up MOX spent fuels with a burn-up of 45 MWd/kgHM was carried out to accumulate nuclide composition data. Furthermore, computational analysis was performed using the integrated burn-up calculation code system SWAT. The differences between the amounts of fission products obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2, and JEFE-3.0 were evaluated as the ratios of the calculated values to the experimental ones (C/E ratios). The fission products such as 88Sr, 90Sr, 106Ru, 133Cs, 134Cs, and 135Cs, which are gamma and decay heat sources, neutron absorption nuclides, or burn-up indicators in spent fuels, were further investigated to improve their C/E ratios using the simplified burn-up chains of fission products using JENDL-3.3; consequently, the correction values for the fission yields or capture cross sections of the fission products were estimated using sensitivity coefficients. The C/E ratios for 154Eu, 155Eu, 154Gd, 155Gd, and 156Gd markedly differed among libraries. The reason for this difference was also discussed using the sensitivity coefficient and capture cross section of each fission product, which is in their main sensitive production paths. (author)

  3. Results of radioactivity measurements in French coastal waters during 1985

    Various sampling networks for the environmental monitoring of levels from radioactive waste releases from french nuclear plants have been set up by the Departement de Protection Sanitaire (DPS) since 1983. In 1985, various marine and freshwater biological indicators were collected regularly on the Channel, Atlantic and Mediterranean shores and at the level of the lower Rhone river. As in the previous years, the results showed the prevailing effects of the releases from the LA HAGUE reprocessing plant on the Channel and of the Rhone waters on the Mediterranean sea. Measurements at the level of the lower Rhone showed a clear labelling of the Rhone river waters by 106Ru. The monitoring of ruthenium levels in various biological indicators from the Mediterranean shore supplied a good representation of the marine areas affected by the Rhone river. At the stations with the highest levels, the sanitary consequence of man-made radionuclides remained low, representing a fraction in the range of 10-5 of the dose limit recommended by the International Commission on Radiological Protection

  4. Migration of technetium-99 in the Nevada Test Site aquifer

    The Hydrology Radionuclide Migration Experiment is measuring the migration of radionuclides from the site of an underground nuclear explosion at the Nevada Test Site. The Cambric event, detonated in 1965, was chosen as the initial experimental site. By 1974, water had returned to its pre-shot level allowing soluble radionuclides to be leached into the water. A re-entry well (RNM-1) was dug into the original cavity to take core and water samples. A satellite well (RNM-2S) was placed 91 meters from the Cambric cavity. Pumping water from this well has induced an artificial gradient which has allowed soluble radionuclides to migrate. Tritium (HTO) has been observed in the RNM-2S water; its elution has been well characterized. Other radionuclides have been monitored in this water: 36Cl, 85Kr, 129I, and 106Ru. The authors have recently measured 99Tc at the 10--20 femtogram/liter level. Technetium appears to be migrating slower than 36Cl and possibly a little faster than tritium

  5. Radioactive substance in the Japanese environment

    The sources of environmental radiation exposure, the features of radioactive substances, the effect of radioactive nuclides to mankind by their intake and so on are explained. The distribution and variation of radioactive substances around Chernobyl Nuclear Power Station and in USSR due to the accident are shown. The worldwide contamination centering around Europe is also shown. In Japan, the rise of radioactivity level due to the accident was small. The radioactive substances which seemed to be originated from the accident were detected from May 3, 1986, and as the nuclides, I-131, Cs-137, Cs-134, Sr-90, Ru-106, Ru-103, Te-132, Ba-140, La-140, Ce-144 and so on were found. Sr-90 somewhat rose in May, but in June, it returned to the state in April, and the level of Cs-137 was somewhat high until July, but returned to the state in April around September. The concentration of radioactive nuclides in foods in Japan was low. The concentration in milk became the maximum in the middle of May, though the maximum fallout occurred on May 3 - 5. As the natural radioactive nuclides in Japan, 110,000 pCi of K-40 and 70,000 kCi of C-14 are contained in a whole body. (Kako, I.)

  6. Thermal stability of the C106 dye in robust electrolytes

    Lund, Torben; Phuong, Nguyen Tuyet; Pechy, Peter; Zakeeruddin, Shaik M.; Grätzel, Michael

    Thermal stability of the C106 dye in robust electrolytes. We have investigated the thermal stability and degradation chemistry of the ruthenium dye C106 (Figure 1) at 80 ◦C in the “robust” electrolyte “B” comprised of 1.0 M DMII, 0.03 M I2, 0.5 M NBB, and 0.1 M GuNCS in 3-methoxypropionitrile (3......-MPN) introduced by Gao et al. in 2008. [1]. Figure 1 Thermal degradation of C106 bound to TiO2 at 80 ºC in dark as a function of heating time. ● C106 = RuLL´(NCS)2 ■ RuLL´(NCS)(NBB)+ ▲ RuLL´(NCS)(3-MPN)+ The C106 dye was attached to the surface of TiO2 nano-particles and stable colloidal solutions of...... the particles were prepared in electrolyte mixture B. The solutions were thermally treated at 80 ◦C for 0-2000 hours followed by dye extraction and analysis by HPLC coupled to UV/Vis and electro spray mass spectrometry [2]. Figure 1 shows the concentration profiles of C106 samples prepared under...

  7. The Cambric migration experiment: A summary report

    The Cambric migration experiment was undertaken in 1974 to provide data concerning the migration of radionuclides away from the site of a nuclear test where there is an induced hydraulic gradient. First, the distribution of radioactive isotopes around the site of the Cambric explosion was investigated by analysis of samples recovered from a hole drilled through the nuclear debris. Then water was pumped from a well, located 91 m away, to induce flow from the nuclear explosion site. Analyses of water samples showed that the migration velocity of tritium, 16Cl, 85Kr, 99Tc, 106Ru, and 129I was nearly the same as that of the moving water, from the explosion site to the pumped well. Less than 0.5% of the total 90Sr and 0.0003% of the total 137Cs accompanied the tritium to the pumped well, although both isotopes appear to have migrated away from the source zone to some extent. The concentration of 239Pu at the pumped well was below the detection limit of 106 atoms/m ell in water collected at the time of peak tritium concentration. Peak tritium concentration in the pumped water occurred when 5,000,000 m3 had been pumped

  8. Performance evaluation of of caesium, iodine, strontium and ruthenium isotopes in urban areas after contamination by accidental release

    The exposures of urban populations to the radiation derived from the deposition, after accidental atmospheric releases, of I37Cs, 134Cs, 129I, 131I, 133I, 89Sr, l03Ru and 106Ru were assessed, using the integrated system for the evaluation of environmental radiological impact in emergency situations (SIEM), developed by the Instituto de Radioprotecao e Dosimetria (IRD) / Comissao Nacional de Energia Nuclear (CNEN). These radionuclides are fission products likely to be emitted in the occurrence of severe nuclear reactor accidents. Their environmental behaviour in urban areas, due to their deposition in soil, in urban surfaces and in vegetable-garden food products, such as leafy and non-leafy vegetables, were analyzed, and dose assessments at the short, medium and long terms were performed, with and without the application of protective measures for reduction of doses. Simulations of unitary initial deposition for each radionuclide and of two different potential accidents involving pressurized water reactors (PWR), with different source terms and distinct deposition for each radionuclide, were performed. Results were analyzed on the basis of the relative relevance of radionuclides and pathways for the exposure of members of the public, as a function of age and time after the release. It was also performed an assessment of the effectiveness of protective measures as a function of the moment of their implementation. (author)

  9. Chromatographic purification of neutron capture molybdenum-99 from cross-contaminant radionuclides

    Technetium-99m is called the work horse, for many reasons, in nuclear medicine diagnostic purposes. It is produced as the β-decay of 99Mo radionuclide. Molybdenum-99 gel type generators are considered as a suitable alternative of the conventional chromatographic alumina columns loaded with fission molybdenum-99. 99Mo neutron-capture is cross-contaminated with radionuclides originated from activation of chemical impurities in the Mo target such 60C0, 65Zn, 95Zr, 175Hf, 181Hf, 86Rb, 134Cs, 141Ce, 152Eu, 140La,51Cr, 124Sb,46Sc, 54Mn, 59Fe and / or fast neutrons interactions with the stable isotopes of molybdenum such as 92mNb, 95Nb and 95Zr. To prevent contamination of the eluted 99mTc, successive purification methods were made. After complete dissolution of the irradiated target wrapped with thin Al foil in 5 M NaOH solution, hydrogen peroxide was added to start precipitation of Fe(OH)3. The formed Fe (III) minerals allow complete elimination of some radio contaminants from the molybdate solute such as 152Eu, 140La,141Ce, 45Mn and 92mNb in addition to partial elimination of 46Sc, 60Co and 59Fe radionuclides. The remaining supernatant was acidified by concentrated nitric acid to ph 9.5 for precipitation of Al(OH)3 with complete elimination of radio contaminants such as 95Zr 175Hf, 181Hf, 65Zn, 124Sb, 51Cr, 46Sc, 60Co and 59Fe. 134Cs and 86Rb radionuclides were not affected by precipitation of Fe(OH)3 or Al(OH)3. Chromatographic column of potassium nickel hexacyanoferrate (II) (KNHCF) has high affinity towards elimination of 134Cs and 86Rb radionuclides. Highly pure molybdate-99Mo solution was processed for preparation of zirconium molybdate gel generator with 99mTc eluate of high radionuclidic, radiochemical and chemical purity suitable for use in medical purposes.

  10. Study on measurement of trace radioactivity, (2)

    The method was worked out so as to measure 59Fe, 60Co, 65Zn, 95Zr, and 144Ce successively from the same sample by ion-exchange separation. This method was also applicable to the measurement of 90Sr and 137Cs. It was confirmed that 131I in milk could be measured satisfactorily by the method of US Atomic Energy Commission, when raw milk before processing was preserved in good condition. 3H in natural water can be measured with a gas magnifying counter without isotope concentration when water is transformed into hydrogen gas. It was found that an external heater type reactor is desirable to transform 3H in water into methane gas so far as the repair is concerned. The present reactor needs some improvement, however, to get better yield as the best synthesizing yield for methane gas obtained so far was only 48.8%. (Kobatake, H.)

  11. Specific processes in solvent extractiotn of radionuclide complexes

    The doctoral thesis discusses the consequences of the radioactive beta transformation in systems liquid-liquid and liquid-ion exchanger, and the effect of the chemical composition of liquid-liquid systems on the distribution of radionuclide traces. A model is derived of radiolysis in two-phase liquid-liquid systems used in nuclear chemical technology. The obtained results are used to suggest the processing of radioactive wastes using the Purex process. For solvent extraction the following radionuclides were used: 59Fe, 95Zr-95Nb, 99Mo, sup(99m)Tc, 99Tc, 103Pd, 137Cs, 141Ce, 144Ce-144Pr, 234Th, and 233Pa. Extraction was carried out at laboratory temperature. 60Co was used as the radiation source. Mainly scintillation spectrometry equipment was used for radiometric analysis. (E.S.)

  12. Development of a cosmic veto gamma-spectrometer

    Cosmic radiation contributes significantly towards the background radiation measured by a gamma-spectrometer. A novel cosmic veto gamma-spectrometer has been developed that provides a mean background reduction of 54.5%. The system consists of plastic scintillation plates operated in time-stamp mode to detect coincident muon interactions within an HPGe gamma-spectrometer. The instrument is easily configurable and provides improved sensitivity for radionuclides indicative of nuclear weapons tests and reactor incidents, including 140Ba, 95Zr, 99Mo, 141Ce, 147Nd, 131I, 134Cs and 137Cs. This has been demonstrated for Comprehensive Nuclear-Test-Ban Treaty applications to obtain the required 140Ba MDA of 24 mBq within 2 days counting. Analysis of an air filter sample collected during the Fukushima incident indicates improved sensitivity compared to conventional gamma-spectrometers. (author)

  13. Induced radioactivities and cross section measurements of the 14 MeV irradiated molybdenum foils

    The radioactivities of 14 MeV neutron irradiated Molybdenum foils have been measured for comparison exercise conducted by the IAEA Nuclear Data Section. The spectra of the characteristic gamma-rays emitted as a result of the induced radioactivity were taken with a Ge(Li) detector and an Intrinsic Germanium detector. The cross sections for the reaction 92Mo(n,np)91mNb, 95Mo(n,p)95Nb and 98Mo(n,α)95Zr have been determined using the information provided by the IAEA on the irradiation time, total fluence and masses of the irradiated foils. The activation cross sections determined from the present measurements have been compared with previous work. (author). 8 refs, figs, 3 tabs

  14. Radionuclides in lake Drukshiai - cooling water reservoir of Ignalina NPP in 1986-1993

    As a result of investigations carried out in 1986-1993 quantitative and qualitative composition of radionuclides in the components of hydroecosystem of lake Drukshiai and their distribution in this lake are presented. Technogenic radionuclides 90Sr, 134Cs, and 137Cs and radionuclides of corrosive origin 54Mn, 60Co, 59Fe, 95Zr, 95Nb, which are characteristic products of NPP action, have been investigated. The interaction of radionuclides, hydrophytes and anthropogenic factors, the sources of radionuclides entering into lake Drukshiai from Ignalina NPP and critical zones of pollution have been established. The intensity of radionuclides accumulation levels in hydrophytes as well as in bottom sediments have been investigated. It has been determined that transformation of radionuclides chemical forms, which take place in lake Drukshiai as a result of the action of various chemical matters, changes physical and chemical properties of radioisotopes and increases their mobility in hydroecosystem.(author). 8 figs., 4 tabs., 6 refs

  15. Neutron capture on (94)Zr: Resonance parameters and Maxwellian-averaged cross sections

    Tagliente, G; Fujii, K; Abbondanno, U; Aerts, G; Alvarez, H; Alvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Audouin, L; Badurek, G; Baumann, P; Becvar, F; Belloni, F; Berthoumieux, E; Bisterzo, S; Calvino, F; Calviani, M; Cano-Ott, D; Capote, R; Carrapico, C; Cennini, P; Chepel, V; Chiaveri, E; Colonna, N; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillmann, I; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Embid-Segura, M; Ferrari, A; Ferreira-Marques, R; Furman, W; Gallino, R; Goncalves, I; Gonzalez-Romero, E; Gramegna, F; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Jericha, E; Kappeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Kossionides, E; Krticka, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Martinez, T; Massimi, C; Mastinu, P; Mengoni, A; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O'Brien, S; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M.T; Plag, R; Plompen, A; Plukis, A; Poch, A; Praena, J; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Santos, C; Sarchiapone, L; Savvidis, I; Stephan, C; Tain, J.L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarin, D; Vincente, M.C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wiescher, M; Wisshak, K

    2011-01-01

    The neutron capture cross sections of the Zr isotopes play an important role in nucleosynthesis studies. The s-process reaction flow between the Fe seed and the heavier isotopes passes through the neutron magic nucleus (90)Zr and through (91,92,93,94)Zr, but only part of the flow extends to (96)Zr because of the branching point at (95)Zr. Apart from their effect on the s-process flow, the comparably small isotopic (n, gamma) cross sections make Zr also an interesting structural material for nuclear reactors. The (94)Zr (n, gamma) cross section has been measured with high resolution at the spallation neutron source n_TOF at CERN and resonance parameters are reported up to 60 keV neutron energy.

  16. Processing of data issued from γ spectrometer. Application to the study of decay schemes

    The main purpose of the following report is the study of a generation method of γ spectra corresponding to any energy, from a library of some basic spectra. Then we point out how to establish the decay scheme of a radio-nuclide, in using an accurate analysis of full-energy peaks which appear in its complex spectrum, followed by a generation of plain spectra corresponding to the energies of the different transitions, and ending with a least-squares fitting procedure of the complex spectrum. The accuracy of the process is examined comparing the calculated spectra to the ones issued from two kinds of spectrometers (Na I(Tl) and Ge(Li)). The method is then applied to the study of 134Cs and 95Zr - 95Nb. (authors)

  17. Contribution to the study of zirconium self-diffusion in zirconium carbide

    The objective of this research thesis is to determine experimental conditions allowing the measurement of the self-diffusion coefficient of zirconium in zirconium carbide. The author reports the development of a method of preparation of zirconium carbide samples. He reports the use of ion implantation as technique to obtain a radio-tracer coating. The obtained results give evidence of the impossibility to use sintered samples with small grains because of the demonstrated importance of intergranular diffusion. The self-diffusion coefficient is obtained in the case of zirconium carbide with grains having a diameter of few millimetres. The presence of 95Nb from the disintegration of 95Zr indicates that these both metallic elements have very close diffusion coefficients at 2.600 C

  18. Measurement of Neutron Activation Cross Sections on Mo isotopes in the Energy Range from 7 MeV to 15 MeV

    Semkova Valentina; Nolte Ralf

    2014-01-01

    An experimental study of the 92Mo(n,p)92Nbm, 92Mo(n,α)89Zr, 95Mo(n,p)95Nbm, 95Mo(n,p)95Nb, 96Mo(n,p)96Nb, 97Mo(n,p)97Nb, 98Mo(n,p)98Nbm, 98Mo(n,a)95Zr, 100Mo(n,α) 97Zr, and 92Mo(n,2n)99Mo activation reaction cross sections were carried out in the 7-15 MeV energy range at the CV28 compact cyclotron at Physikalisch-Technische Bundesanstalt, Braunschweig. The PTB TOF spectrometer with a D(d,n) source is well suited for this difficult energy range were significant correction for non-monoenergetic...

  19. Determination of hafnium and zirconium in geological materials by neutron activation analysis

    In this paper, neutron activation analysis was developed for determining hafnium and zirconium in geological materials. The USGS geological standard rocks GSP-1 (granodiorite) and W-1 (di abase). The Brazilian geological standards GB-1 (granite) and BB-1 (basalt) from Instituto de Geociencias da Universidade da Bahia and P-1 a uraniferous rock from Pocos de Caldas, MG, Brazil were analyzed. Hafnium present in these rocks was analyzed by purely instrumental method by irradiating with both thermal and epithermal neutrons from IEA-R1 nuclear research reactor. In the case of zirconium depending on the sample a radiochemical separation was required. 154 Eu and 152 Eu radioisotopes emit gamma rays with energies too close to those emitted by 95 Zr and they cause interferences. (author)

  20. Results of examination of the personnel at Novo Voronezh nuclear power plant using human radiation counter

    Results of spot examination of the personnel of Novo- Voronezh NPP (56 persons) in 1976 by a portable human radiation counter are given. Measurements were performed with a device consisting of two NaI(Tl) scintillation gamma detectors and LP4840 type multichannel amplitude analyzer. One of the detectors recorded thyroid gland radiation in the range of 0.03-1.06 MeV, another one recorded radiation from chest (lungs, gastroenteric tract) in the range of 0.06-2.09 MeV. It is shown that intake of 60Co, 54Mn, 58Co, 95Zr, 95Nb, 53Co, 134Cs, 131I radionuc= . lides and their accumulation in thyroid gland and chest does not exceed hundredth or tens of shares of permissible concentration recommended in radiation safety standards, 1976 (RSS-76)

  1. EML Surface Air Sampling Program, 1990--1993 data

    Measurements of the concentrations of specific atmospheric radionuclides in air filter samples collected for the Environmental Measurements Laboratory's Surface Air Sampling Program (SASP) during 1990--1993, with the exception of April 1993, indicate that anthropogenic radionuclides, in both hemispheres, were at or below the lower limits of detection for the sampling and analytical techniques that were used to collect and measure them. The occasional detection of 137Cs in some air filter samples may have resulted from resuspension of previously deposited debris. Following the April 6, 1993 accident and release of radionuclides into the atmosphere at a reprocessing plant in the Tomsk-7 military nuclear complex located 16 km north of the Siberian city of Tomsk, Russia, weekly air filter samples from Barrow, Alaska; Thule, Greenland and Moosonee, Canada were selected for special analyses. The naturally occurring radioisotopes that the authors measure, 7Be and 210Pb, continue to be detected in most air filter samples. Variations in the annual mean concentrations of 7Be at many of the sites appear to result primarily from changes in the atmospheric production rate of this cosmogenic radionuclide. Short-term variations in the concentrations of 7Be and 210Pb continued to be observed at many sites at which weekly air filter samples were analyzed. The monthly gross gamma-ray activity and the monthly mean surface air concentrations of 7Be, 95Zr, 137Cs, 144Ce, and 210Pb measured at sampling sites in SASP during 1990--1993 are presented. The weekly mean surface air concentrations of 7Be, 95Zr, 137Cs, 144Ce, and 210Pb for samples collected during 1990--1993 are given for 17 sites

  2. A selective separation method for 93Zr in radiochemical analysis of low and intermediate level wastes from nuclear power plants

    The zirconium isotope 93Zr is a long-lived pure β-particle-emitting radionuclide produced from 235U fission and from neutron activation of the stable isotope 92Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, 93Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of 93Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. A radiochemical procedure based on liquid-liquid extraction with 1-(2-thenoyl)-3,3,3-trifluoroacetone in xylene, ion exchange with Dowex resin and selective extraction using TRU resin has to be carried out in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique (LSC). To set up the radiochemical separation procedure for 93Zr, a tracer solution of 95Zr was used in order to follow the behavior of zirconium during the process by γ-ray spectrometry through measurement of the 95Zr. Then, the protocol was applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. The efficiency detection for 63Ni was used to determination of 93Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq l-1 was obtained for 63Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for OptiPhase HiSafe 3 cocktail. (author)

  3. EML Surface Air Sampling Program, 1990--1993 data

    Larsen, R.J.; Sanderson, C.G.; Kada, J.

    1995-11-01

    Measurements of the concentrations of specific atmospheric radionuclides in air filter samples collected for the Environmental Measurements Laboratory`s Surface Air Sampling Program (SASP) during 1990--1993, with the exception of April 1993, indicate that anthropogenic radionuclides, in both hemispheres, were at or below the lower limits of detection for the sampling and analytical techniques that were used to collect and measure them. The occasional detection of {sup 137}Cs in some air filter samples may have resulted from resuspension of previously deposited debris. Following the April 6, 1993 accident and release of radionuclides into the atmosphere at a reprocessing plant in the Tomsk-7 military nuclear complex located 16 km north of the Siberian city of Tomsk, Russia, weekly air filter samples from Barrow, Alaska; Thule, Greenland and Moosonee, Canada were selected for special analyses. The naturally occurring radioisotopes that the authors measure, {sup 7}Be and {sup 210}Pb, continue to be detected in most air filter samples. Variations in the annual mean concentrations of {sup 7}Be at many of the sites appear to result primarily from changes in the atmospheric production rate of this cosmogenic radionuclide. Short-term variations in the concentrations of {sup 7}Be and {sup 210}Pb continued to be observed at many sites at which weekly air filter samples were analyzed. The monthly gross gamma-ray activity and the monthly mean surface air concentrations of {sup 7}Be, {sup 95}Zr, {sup 137}Cs, {sup 144}Ce, and {sup 210}Pb measured at sampling sites in SASP during 1990--1993 are presented. The weekly mean surface air concentrations of {sup 7}Be, {sup 95}Zr, {sup 137}Cs, {sup 144}Ce, and {sup 210}Pb for samples collected during 1990--1993 are given for 17 sites.

  4. Fetoplacental transport of various trace elements in pregnant rat using the multitracer technique

    The placenta functions as the barrier between fetus and mother, providing means of regulation of heat exchange, respiration, nutrition, and excretion for the fetus. In this paper, the multitracer technique was applied to study the maternal transport of trace elements via the placenta to the fetus. In this experiment, the multitracer solution used contained the following nuclides: 7Be, 22Na, 46Sc, 48V, 52Mn, 59Fe, 56Co, 65Zn, 67Ga, 74As, 75Se, 84Rb, 85Sr, 87Y, 88Zr, 96Tc, 101mRh, and 103Ru. We examined the time dependence of the uptake amounts about various elements. From these results, we observed a large difference in the time dependencies between elements and the elements were classified into three groups. Group I elements, such as Be, Sc, V, As, Y, Zr, Tc, Rh, and Ru, are transported to the placenta from the maternal blood and only accumulates in the placenta. Group II elements, such as Na, Co, Ga, Rb, and Sr, are transported to the placenta from the maternal blood and accumulate in the placenta, fetus, and amniotic fluid. Group III elements, such as Mn, Fe, Zn, and Se, are transported to the placenta from the maternal blood and mainly accumulate in the fetus. From these results, it was considered that the placenta is a highly selective filters because essential elements such as Group III elements are readily transported from the placental membrane to the growing fetus, whereas nonessential metals such as Group I elements have difficulty penetrating the placental barrier that protects the fetus from the toxic effects of these elements. (author)

  5. Forage particle breakdown and movement in the reticulo-rumen in cattle and interactions between cold exposure and diets in sheep

    Sheep, fitted with cannulas in the rumen and proximal duodenum, were given different diets as well as different physical forms of diets. The sheep were closely shorn and exposed to ambient temperatures of 22 deg. C to 24 deg. C (warm) or 1 deg. C to 5 deg. C (cold). Measurements of rates of passage, digestibility, rate of transfer of plasma urea to the rumen, feed protein degradation, microbial protein synthesis and rate of rumen fermentation were determined. The effect of cold exposure on digestion was not dependent on the physical form of the diet given at a fixed intake, but was associated with the type of diet (hay, concentrate). The greater voluntary intake was a result of increased clearance of digesta from the rumen through decreased particle size and increased frequency of reticulum contractions. The apparent digestibility of organic matter and nitrogen was reduced for hay diets in cold exposure, which increased the escape of dietary N from the abomasum and the efficiency of microbial synthesis. Cold exposure did not affect urea transfer from plasma to the rumen. Subsequently, two steers fitted with cannulas in the oesophagus and rumen were used in a 2x2 cross-over design to study the physical breakdown of different diets. Concomitantly, four steers, each fitted with a rumen cannula, were used in a 4x4 Latin square design to investigate removal of digesta components from the rumen. A combination of wet sieving and marker techniques for fluid, particulate and microbial movements was utilized. The fractional outflow rate (FOR) of 51Cr-EDTA in centrifuged rumen fluid was increased (P103Ru-P in mixed rumen digesta and organic 35S in microorganisms were linearly correlated (P0.05) by grinding and salt treatments

  6. Amniotic fluid volume and fetal swallowing rate in sheep

    To investigate amniotic fluid (AF) dynamics and volume regulatory mechanisms, the authors measured the concentration of radioiodinated (125I) serum albumin (RISA), 51Cr-labeled red blood cells (Cr-RBC), and 103Ru-labeled microspheres after injection into the amniotic cavity and determined AF volume and fetal swallowing rate in 22 singleton pregnant sheep. Under normal conditions 2-3 h were required for complete mixing of RISA and Cr-RBC within AF; however, when the fetus was dead only 3-5 h were required. AF volume of 17 sheep on the 5th postoperative day averaged 975 +/- 128 ml by RISA and 986 +/- 130 ml by Cr-RBC. AF volume determined with RISA and Cr-RBC correlated well. In contrast, AF volume measurement with microspheres produced erratic results. The disappearance rate of the labels in 17 ewes on the 5th postoperative day averaged 4.9 +/- 0.7%/h for RISA and 5.5 +/- 0.7 for Cr-RBC, and the calculated rates of fetal swallowing were 935 +/- 78 ml/day by RISA and 1,085 +/- 102 by Cr-RBC. In dead fetuses the disappearance rates were almost zero, suggesting that the labels disappear mainly by swallowing. Absolute volume swallowed and swallowed volume per fetal weight correlated with gestational age. AF volume correlated with fetal weight. Radiolabeled albumin or red blood cells may be used to simultaneously measure amniotic fluid volume and the rate of fetal swallowing. Furthermore it appears that fetal swallowing increases with gestational age

  7. Selection of polymer solid supports for supported liquid membrane applications

    The role of polymer support selection upon the performance and stability of a supported liquid membrane(SLM) has been evaluated in detail. Permeation experiments are carried out batch wise in a simple twin stirred cells. Several solid supports such as polytetrafluroethylene (TE-35, TE-36, TE-37), polypropylene (HF-PP), cellulose nitrate (BA-S-83) of different pore sizes and varying thickness are tested. Dicyclohexano-18-crown-6 (DC18C6) in toluene has been employed as mobile carrier in permeation of plutonium (IV) nitrate across SLM. Selectivity of plutonium transport from fission product contaminants such as 137Cs. 103Ru and 144Ce is found to be lesser with large pore size (0.45 μm) support. Chemical stability of these hydrophobic polymer membranes against organic solvents and nitric acid is evaluated. Polytetrafluroethylene supports like TE-35, TE-36, TE-37 and polypropylene HF-PP membranes have not shown any marked deterioration but cellulose nitrate membrane such as BA-S-83 is attacked by organic solvent. Such deleterious effects have been confirmed by the scanning electron microscopy which revealed that pore structure becomes completely blocked. Radiolytic stability of these polymers is also tested by irradiating them up to 3 Mrads using 60Co irradiator. No serious damage due to radiation is observed as micrographs have shown undisturbed network of polymer structure even after the irradiation. Of the different polymer supports tested, both TE-35 and HF-PP thin-film solid membranes prove to be quite suitable for SLM purposes as their optimum pore size, good physical and chemical stability including radiation stability render them relatively more useful for varied applications in liquid membranes. (author). 5 refs., 3 figs., 3 tabs

  8. Advanced orient cycle, for strategic separation, transmutation and utilization of nuclides in the nuclear fuel cycle

    directly recover pure Cm as well as pure Am with minimum number of reprocessing separation steps is reported in another paper. The recent experiments indicated that strong adsorption of 106Ru and 125Sb was observed under the diluted HCl medium, thereby completely 106Ru-free feed dissolver solution was obtained. The CEE separation step will follow this IX step for further purification and fabrication of RMFP material for their utilization. Based on those technologies, the Trinitarian Research and Development project (Advanced ORIENT Cycle) on partitioning, transmutation and utilization of actinides and fission products will be developed to realize ultimate reducing long-term radio toxicity in the radioactive wastes. Actinides, LLFP (135Cs, etc), MLFP (90Sr, 137Cs) and RMFP shall be separated to the level of isotope as well as element. The CEE process will be added for utilization of RMFP. The RMFP, one of the products of Ad. ORIENT Cycle, would be expected to be a 'FP-catalyst' to circulate between nuclear and hydrogen / fuel cell energy systems, and thereby contributing to save the natural precious metal resources

  9. Department of Radiation Shielding and Dosimetry: Overview

    Full text: The research activities of the Department in 1998, similarly to the previous year were focused on the following problems: Dosimetry for medical purposes; Microdosimetry at the nanometer level; Numerical modelling of interaction of radiation with matter; DOSIMETRY: Based on experience gained in previous years in absolute and relative measurements of absorbed dose for 106Ru applicators, the detectors and methods for dosimetry of β radiation applied in intravascular brachytherapy have been undertaken. A new, small size scintillation probe with NE102A scintillator 1 mm dia. by 1 mm coupled to a 30 cm long flexible light guide and to a 9524S photomultiplier has been assembled and tested. The GAF Chromic foils, MDSS, have been found to be very promising detectors for intravascular and ocular brachytherapy. A miniature ionisation chamber for Kerma in air measurements in radiation field of a ''photon needle'' (small size X-ray tube operated at 30 KV) has been assembled and tested. MICRODOSIMETRY: The absolute efficiency of two types of electron multipliers, i.e. discrete dynode electron multiplier DM205IG and channel electron multiplier X719BL for Ar+ ions in energy range 1 keV to 10 keV has been determined in an experiment performed in cooperation with the Weizmann Institute of Science. These electron multipliers are used in the set up ''JET COUNTER'' as detectors for ion cluster studies. A method for measuring the spectra of ion clusters created along a charged particles track has been proposed. The ion clusters spectra produced by alpha particles 241Am source passed a distance of 3.6 to 10 nm (in units of density scale) in nitrogen have been measured. Also, preliminary measurements of ion clusters created by low energy electrons 50 and 100 eV have been carried out. Activities in this field were supported by IV CEC Framework Programme as well as by the Polish State Commission for Scientific Research. NUMERICAL MODELLING: Monte Carlo simulation is direct and

  10. Speciation and transport of radionuclides in ground water

    Studies of the chemical speciation of a number of radionuclides migrating in a slightly contaminated ground water plume are identifying the most mobile species and providing an opportunity to test and/or validate geochemical models of radionuclide transport in ground waters. Results to date have shown that most of the migrating radionuclides are present in anionic or nonionic forms. These include anionic forms of 55Fe, 60Co, /sup 99m/Tc, 106Ru, 131I, and nonionic forms of 63Ni and 125Sb. Strontium-70 and a small fraction of the mobile 60Co are the only cationic radionuclides which have been detected moving in the ground water plume beyond 30 meters from the source. A comparison of the observed chemical forms with the predicted species calculated from modeling thermodynamic data and ground water chemical parameters has indicated a good agreement for most of the radioelements in the system, including Tc, Np, Cs, Sr, Ce, Ru, Sb, Zn, and Mn. The discrepancies between observed and calculated solutions species were noted for Fe, Co, Ni and I. Traces of Fe, Co, and Ni were observed to migrate in anionic or nonionic forms which the calculations failed to predict. These anionic/nonionic species may be organic complexes having enhanced mobility in ground waters. The radioiodine, for example, was shown to behave totally as an anion but further investigation revealed that 49-57% of this anionic iodine was organically bound. The ground water and aqueous extracts of trench sediments contain a wide variety of organic compounds, some of which could serve as complexing agents for the radionuclides. These results indicate the need for further research at a variety of field sites in defining precisely the chemical forms of the mobile radionuclide species, and in better understanding the role of dissolved organic materials in ground water transport of radionuclides

  11. Results of the Separations Area ground-water monitoring network for 1981

    This report summarizes and interprets the results of the Separations Area ground-water monitoring program for calendar year 1981. The Separations Area ground-water monitoring program satisfies US Department of Energy requirements that all onsite discharges be monitored. There are 105 water quality monitoring wells in the 1981 Separations Area monitoring network. Samples from the water quality monitoring wells were collected at monthly or semiannual intervals. These samples were analyzed selectively for total alpha, total beta, 60Co, 106Ru, 137Cs, 90Sr, tritium, uranium, and nitrate. Review of 1981 results indicated few changes from 1980. The 3000-pCi/mL tritium guideline was exceeded in three wells that monitor two active sites: Well 299-E24-2 monitors the 216-A-10 Crib, and Wells 299-E25-18 and 299-E25-20 monitor 219-A-37 Crib. Sample concentrations reflect plume migrations of active sites. With the exception of tritium, all contamination greater than 10% of Table II guidelines in the ground water under the Separations Area was caused by past disposal to inactive sites. The only samples with contamination that exceeded Table II guidelines were collected from Wells 299-W22-1, 299-E28-23, and 299-E33-27, which monitor the inactive 216-S-1 and -2 Crib, 216-B-5 Reverse Well, and 216-BX Tank Farm. However, the zone of contamination appears stable and highly localized for these sites. Analysis of contamination plumes within the Separations Area indicates that contamination levels in ground water migrating out of the Separations Area at this time are less than the DOE Table II guidelines

  12. Passive gamma analysis of the boiling-water-reactor assemblies

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  13. Status report on LWR spent fuel IAEA leach tests

    Spent light-water-reactor (LWR) fuel with an average burnup of 28,000 MWd/MTU was leach-tested at 250C using a modified version of the International Atomic Energy Agency procedure. Leach rates were determined from tests conducted in five different solutions: deionized water, sodium chloride (NaCl), sodium bicarbonate (NaHCO3), calcium chloride (CaCl2) and Waste Isolation Pilot Plant B brine solutions. Elemental leach rates are reported based on the release of 90Sr + 90Y, 106Ru, 137Cs, 144Ce, 154Eu, /sup 239 + 240/Pu, 244Cm and total uranium. After 467 days of cumulative leaching, the elemental leach rates are highest in deionized water. The elemental leach rates uin the different solutions generally decreased from deionized water to the 0.03M NaCl solution to the WIPP B brine solution to the 0.03M NaHCO3 solution and was a factor of 20 lower in 0.015M CaCl2 solution than in deionized water. The leach rates of spent fuel and borosilicate waste-glass were also compared. In sodium bicarbonate solution, the leach rates of the two waste forms were nearly equal, but the glass was increasingly more resistant than spent fuel in calcium chloride solution, followed by sodium chloride solution, WIPP B brine solution and deionized water. In deionized water the glass, based on the elemental release of plutonium and curium, was 50 to 400 times more leach resistant than spent fuel

  14. Preparation, characterisation and evaluation of nano MnO2 for uptake of alpha contaminants from radioactive liquid waste

    Radioactive liquid waste is generated during reprocessing of spent nuclear fuel, which is classified as, low level liquid waste (LLW), intermediate level waste (ILW) and high level waste (HLW) based on radioactivity concentration. The strategy for management of LLW and ILW lies in the development and deployment of innovative processes having volume reduction as one of the important parameter. For achieving volume reduction, methodologies based on separation science such as ion exchange, adsorption, extraction etc. are used. Effluents generated during such treatment of LLW will have small concentration of radionuclides like 137Cs, 90Sr, 106Ru, 238U, 239Pu, 241Am etc. These waste streams require treatment to further reduce their activity content following near zero discharge approach. Nano-materials based on the mineral oxides of Si, Al, Ti, Mn, etc., are used as sorbents for heavy metal ions present in waste waters. Oxidative nature of MnO2 has been shown to regulate the concentration of some of the toxic metal ions, viz., Cr, Pb, in estuarine waters. Keeping these properties of MnO2 in mind, in the present study, nanocrystalline manganese oxide was prepared by hydrolysis of potassium permanganate by ethanol. A FT-IR spectrum of the powder was recorded to ensure that there is no organic impurity in the sample due to the alcoholic hydrolysis process. The oxide obtained was characterized for phase analysis and crystallite size estimation using X-ray diffraction technique. The average crystallite size calculated from the Scherrer's formula was found to be ∼ 4 to 8 nm, which is further confirmed from the TEM micrographs. The nano MnO2was evaluated for the sorption characteristics of Pu

  15. The potentialities of the complexation ultrafiltration technique for the decontamination of fission product contaminated aqueous effluents

    Many nuclear researchers and industrial operators lay emphasis on improving the back end of the fuel cycle. A major problem concerns the liquid wastes generated by the reprocessing plant at La Hague, discharged into the sea after treatment in the Effluent Treatment Station (STE) 3), and which have become crucial matter. The activity of these wastes is well below the current legal limits, and is constantly decreasing these last years. To bring it close to zero, and ambitious goal, entails innovative new reprocessing techniques. We accordingly investigated the possibilities of complexation-ultrafiltration, a technique that uses water-soluble macromolecules to complex the target elements to be separated. We first achieved the strontium (II) separation with poly-acrylic and poly-sulfonic acids. The effects of pH and NaNO3 concentration influence on Sr (II) complexation were studied. The Sr (II) complexation and concentration phases, followed by cation de-complexation to recover the polymer, were also taken into account. This research, combined with a potentiometric study of the polymers, offered a close understanding of the chemical systems involved, and of the operating conditions and limits of complexation-ultrafiltration. The laboratory results were also validated on a tangential ultrafiltration pilot plant. We then used complexation-ultrafiltration to treat a real effluent generated bu La Hague's STE 3 plant. This experiment demonstrated minimum 90 % decontamination of Sr (II) (with polyacrylate complexing agent), and also for 134-137Cs (with simple ultrafiltration). The use of two polyamides allowed partial decontamination of the effluent for 60Co and 106Ru. This work therefore offers a global approach to complexation-ultrafiltration, from laboratory to pilot scale, on real and simulated effluents. The future of this technique relies chiefly on the ability to solve the problem of polymer recovery. In other respect, complexation-ultrafiltration clearly offers a

  16. Present Status of the JENDL Project (May, 2014)

    The Japanese Nuclear Data Committee (JNDC) is a research committee for JAEA research activities. It has two subcommittees: Subcommittee on Nuclear Data and Subcommittee on Reactor Constants. The JENDL-4.0 Updated Files (JENDL-4.0u) have been produced to take care of errors which were found after the release of JENDL-4.0. The JENDL-4.0u includes the nuclides whose nuclear data partly revised from important and/or trivial error(s). Thirteen files for JENDL-4.0 were released in 2013. Benchmarking for fission reactor applications was done using test files of Gd-157 and Am-241, and an integral testing for Th-232 was done using KUCA critical experiments. The data which are not updated in JENDL-4.0 are continuously considered to be revised, especially for FP region nuclides. Evaluated results for 141,143Pr, 121,123-126Sb and 96,98-106Ru were published. The R-matrix analysis for O-16 is in progress. For decommissioning of nuclear power plants, evaluation of activation cross sections was planned for 569 reactions with 309 nuclides and is in progress. Evaluated files for the 246 isotopes from H to Hf have been created. As the activity of High Energy Nuclear Data Evaluation WG, development of new JENDL Photonuclear Data File is in progress. Evaluation for 181 nuclides has been finished. A Symposium on Nuclear Data was held on 14-15 November 2013 at Research Institute of Nuclear Engineering, University of Fukui, Tsuruga-shi, with topics about Progress in neutron cross-section measurement and analysis, Application of Nuclear Data, Recent topics, and Progress in studies of high-energy nuclear reaction. A total of 37 papers including 23 poster presentations were presented

  17. Subsurface radionuclide investigation of a nuclear test

    Mathews, M.; Hahn, K.; Thompson, J.; Gadeken, L.; Madigan, W.

    1994-08-01

    This paper reports on an environmental investigation into the vertical distribution of radionuclides from a nuclear test. Dalhart is the name of an underground nuclear test that was executed at the Nevada Test Site at a depth of 2100 ft on October 13, 1988. The test occurred below the static water level of 1667 ft and created multiple radioactive isotopes or fission products. These radioactive isotopes penetrated the surrounding formations and chimney region above the test and were retained there. A 19° 9- {7}/{8}-inch diameter slant hole was drilled to sample the geologic material in the chimney region above the Dalhart test for the purpose of assessing the distribution of radioactivity in and around the shot site. A 30-ft core recovered from a vertical depth of 1628 ft in the collapsed zone or chimney region and above the original static water level was found to be free of radionuclides. Drilling was completed to a vertical depth of 2156 ft with the present static water level at a vertical depth of 1644 ft. Gamma-ray spectroscopy log measurements, made within the drill pipe while drilling fluid was pumped through this pipe, indicate that radioactive material produced by the test was present from the vertical depth interval of 1746-2156 ft. Side-wall samples acquired from the vertical depth interval of 1721-2089 ft and analyzed in the field contained radionuclides such as 137Cs, 125Sb, 106Ru, plus the natural radioactive background of potassium, uranium, and thorium. These samples were sent to Los Alamos to determine the complete radionuclide content at each depth. These analyses were used with the gamma-ray spectroscopy logging data to determine the subsurface vertical radionuclide distribution at the Dalhart site.

  18. Subsurface radionuclide investigation of a nuclear test

    This paper reports on an environmental investigation into the vertical distribution of radionuclides from a nuclear test. Dalhart is the name of an underground nuclear test that was executed at the Nevada Test Site at a depth of 2100 ft on October 13, 1988. The test occurred below the static water level of 1667 ft and created multiple radioactive isotopes or fission products. These radioactive isotopes penetrated the surrounding formations and chimney region above the test and were retained there. A 19o 9-7/8-inch diameter slant hole was drilled to sample the geologic material in the chimney region above the Dalhart test for the purpose of assessing the distribution of radioactivity in and around the shot site. A 30-ft core recovered from a vertical depth of 1628 ft in the collapsed zone or chimney region and above the original static water level was found to be free of radionuclides. Drilling was completed to a vertical depth of 2156 ft with the present static water level at a vertical depth of 1644 ft. Gamma-ray spectroscopy log measurements, made within the drill pipe while drilling fluid was pumped through this pipe, indicate that radioactive material produced by the test was present from the vertical depth interval of 1746-2156 ft. Side-wall samples acquired from the vertical depth interval of 1721-2089 ft and analyzed in the field contained radionuclides such as 137Cs, 125Sb, 106Ru, plus the natural radioactive background of potassium, uranium, and thorium. These samples were sent to Los Alamos to determine the complete radionuclide content at each depth. These analyses were used with the gamma-ray spectroscopy logging data to determine the subsurface vertical radionuclide distribution at the Dalhart site

  19. Experimental study on biokinetics of radionuclides in age groups

    In the aftermath of the Chernobyl accident, it becomes evident that dose coefficients for members of the public are necessary. International Commission on Radiological Protection (ICRP) established a task group of Committee 2 charged with the assessment of dose coefficients as a function of an individual's age. However, little data is available on the biokinetics of radionuclides in juvenile and is a need to develop age-dependent biokinetic models, such as for the gastrointestinal tract. The present paper reviewed an outline on characteristics of biokinetics of radionuclides in juvenile animals focusing on the previous experimental data. The following radionuclides are discussed: 54Mn, 60Co, 65Zn, 75Se, 106Ru, 110mAg, 115Cd, 125Sb, 137Cs, 141Ce, 203Hg and 3H. Generally, intestinal absorption and whole-body retention of radionuclides in Juveniles were higher than that of adult. In the case of sucklings, it is very important to study how radionuclides are transferred through the placenta and milk. The transfer rate of radionuclides through the placenta and milk is dependent on the period of gestation at the time of dosing. The IDES (Internal Dose Estimation System) which is based on the ICRP model was used for dose calculation. We modified the IDES using the biokinetic data which was gained by animal experiment. The IDES is flexible because the absorbed dose can be calculated by substituting arbitrary physical and physiological parameters and also substituting ingested dose coefficients not only for the ICRP Reference Man, but also for Japanese of 1 year old, 5 years old, 10 years old, 15 years old and the adult, respectively. (author)

  20. Transfer of nuclides from the water phase to the sediments during normal and extraordinary hydrological cycles

    Atucha I and Atucha II nuclear power plants are located on the right margin of the Parana de las Palmas river. This river belongs to the Cuenca del Plata, whose 1982-1983 hydrologic cycle registered the greatest freshets of the century. Works and studies previously fixed had to be altered and investigations were adapted to the possibilities and the particular hydric conditions verified. Considerations on the transfer of nuclides between water and sediments are presented. The floods reduce the water-sediments contact time on the bed of the river. In outer areas, the waters labelled by the nuclear power plant effluent discharge favor the infiltration in alluvial soils, as well as the exchange with the sediments. The investigations carried out for the phase near to the discharge of liquid effluents (related to the critical group) made possible to prove the characteristics of the path of the liquid wastes released, the distribution coefficient and the fixation or penetrability of some nuclides in soils of the floody valley. In this manner, a balance of radioactive nuclides incorporated to soils and sediments from the neighbourhood of Atucha and the water-course of Parana de las Palmas river is obtained. The presence of 60Co and 137Cs in the floody soils on the right margin of this river was detected and measured during the greatest flood of the century. On the other hand, 144Ce, 51Cr, 106Ru and 90Sr have not been detected. The detection of artificial radioisotopes turns out to be impossible in normal hydrological years, even in the sorroundings of the nuclear power plant or the critical group (from the point of view of the surface waters, The Fishing Club, 3 km down stream). (M.E.L.)

  1. Estimation of 90Sr activity in reprocessed uranium from the PUREX process

    90Sr estimation in reprocessed uranium was carried out by a series of solvent extraction and carrier precipitation techniques using strontium and lanthanum carriers. Fuming with HClO4 was used to remove 106Ru as RuO4. Three step solvent extraction with 50% tri-n-butyl phosphate in xylene in presence of small amounts of dibutyl phosphate and thenoyl trifluoro acetone was carried out to eliminate uranium, plutonium, thorium and protactinium impurities. Lanthanum oxalate precipitation in acid medium was employed to scavenge the remaining multivalent ions. Strontium was precipitated as strontium oxalate in alkaline pH and 137Cs was removed by washing the precipitate with water. A strontium recovery well above 70% was obtained. Final estimation was carried out by radiometry using end window GM counter after drying the precipitate under an infra red lamp. The same procedure was extended to the estimation of 90Sr in a diluted sample of the actual spent fuel solution. An additional lanthanum oxalate precipitation step was required to remove the entire 144Ce impurity from this sample. This modified procedure was employed in the determination of 90Sr in a number of reprocessed uranium samples and the over all precision of the method was found to be well within ±10%. An additional barium chromate precipitation step was necessary for the analysis of reprocessed uranium samples from high burnup fuels to eliminate trace amounts of short lived 224Ra produced during the decay of 232U and its daughters as they interfere in the estimation of 90Sr. (author)

  2. Urban radionuclide contamination studied in sewage water and sludge from Lund and Gothenburg

    The concentrations of various radionuclides have been measured in the sewage treatment plant serving the town of Lund in southern Sweden. The incoming water, the outgoing water and the sludge have been studied. The measurements started already in 1983 and have gone on since then, which means both before and after the Chernobyl accident. After the accident the concentration of radionuclides was also measured at Gothenburg in western Sweden. The total deposition of the various radionuclides was almost the same 2 kBq/m2 at Lund and Gothenburg although the temporal distribution was somewhat different. The activity concentration measured in the sludge reached a peak value of 87.1 kBq/kg (d.w) for 106Ru and 3.8 kBq/kg for 137Cs at Gothenburgh and of 15.4 and 12.1 k/Bq/kg respectively at Lund. These activity concentrations were measured about 2-3 days after at Lund. The differences in the delays are due to differences in the construction of the plants. The close investigation of the sewage treatment plant at Lund makes it possible to state that about 50 per cent of the incoming 137Cs leaves the plant with the outgoing sludge. It is reasonable to believe that 50 per cent also is valid for the plant at Gothenburg. At Lund the total yearly outgoing activity of 137Cs has fallen exponentially from 900 MBq in 1986 to 100 MBq in 1989. An extrapolation reveals that the activity concentration will be back to pre-Chernobyl levels in 1992/93. But of the original deposition of 12.400 MBq 2/3 or 8.200 MBq will still be left. The same retention factor most probably goes for Gothenburg too. (authors)

  3. Field and model investigations of external gamma dose rates along the Cumbrian coast, NW England

    McDonald, P [Environmental Sciences, Westlakes Scientific Consulting Limited, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3LN (United Kingdom); Bryan, S E [Environmental Sciences, Westlakes Scientific Consulting Limited, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3LN (United Kingdom); Hunt, G J [Centre for Environment, Fisheries and Aquaculture Science, CEFAS Lowestoft Laboratory, Pakefield Road, Lowestoft, Suffolk NR33 0HT (United Kingdom); Baldwin, M [Centre for Environment, Fisheries and Aquaculture Science, CEFAS Lowestoft Laboratory, Pakefield Road, Lowestoft, Suffolk NR33 0HT (United Kingdom); Parker, T G [British Nuclear Group, Sellafield, Seascale, Cumbria CA20 1PG (United Kingdom)

    2005-03-01

    A survey of the contribution to external dose from gamma rays originating from intertidal sediments in the vicinity of the British Nuclear Group Sellafield site showed that the major anthropogenic contributions were due to {sup 137}Cs and {sup 60}Co. At some sites, traces of other anthropogenic radionuclides were detected, namely {sup 106}Ru, {sup 125}Sb, and {sup 154}Eu. The proportions of fine grained material (<63 {mu}m) were used to improve model predictions of dose contribution due to external exposure to gamma rays, using the CUMBRIA77/DOSE77 model. Model dose predictions were compared to those directly measured in the field. Using the new proportions of fine grained material (1-17.5%) in conjunction with field gamma-ray spectra, model predictions were improved considerably for most sites. Exceptions were at Drigg Barn Scar and Whitehaven Coal Sands sites, which had their own unique characteristics. The highest {sup 60}Co activity concentrations in this study were detected at Drigg Barn Scar. These relatively high activity concentrations of {sup 60}Co were due to the presence of {sup 60}Co in mussels and barnacles, hence upsetting the fine sediment relationships used in previous dose calculations. Whitehaven Coal Sands was unusual in that it contained higher levels of radionuclides than would be expected in sandy sediment. The mineralogy of these sediments was the controlling factor on {sup 137}Cs binding, rather than the proportion of fine grained material. By adjusting the effective fine grained sediment proportions for calculations involving {sup 60}Co and {sup 137}Cs at Drigg Barn Scar and Whitehaven Coal Sands respectively, the CUMBRIA77/DOSE77 model predictions could be improved upon significantly for these sites. This work highlights the influence of particle size and sediment composition on external dose rate calculations, as well as the potential for external dose contributions from biota.

  4. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  5. Sulphur containing novel extractants for extraction-separation of palladium (II)

    Extraction performance of palladium (II) by sulphur containing extragents has unequivocally established their strong extraction ability toward this thiophilic soft metal. Hence a comprehensive investigative study was initiated by us to examine selective reversible extraction-separation of trace and macro amounts of palladium (II) from both aqueous nitric acid as well as hydrochloric acid media into 1,2-dichloroethane by 1,10-dithia-18 crown-6 (1,10-DT18C6), S6-pentano-36 (S6-P-36) and bis (2-ethylhexyl) sulphoxide (BESO) dissolved in toluene. From the study of aqueous phase acidity, reagent concentration, period of equilibration, diluent, strippant and diverse ions, conditions are established from its quantitative and reversible extraction. Recovery of Pd(II) from loaded thiacrown and sulphoxide phase is easily accomplished by using sodium thiocyanate, ammonium thiocyanate, thiourea, sodium thiosulphate and mixture of (2M Na2CO3 + 0.5 NH4OH) (only for BESO) as the strippants. The lack of interference from even appreciable amounts of contaminants like 137Cs, 106Ru, 233U and 239Pu may be considered as one of the outstanding advantages of the method. Application of these extractants has been successfully tested for the recovery of palladium from high active waste matrix. The extracted complex from both the thiacrowns has been characterized by elemental analyses and UV-Visible spectra, confirmed to be PdA2.T (A = NO-3, Cl-) from dilute (pH ∼ 2) acid solutions while composition of organic species with palladium for the sulphoxide, has also been confirmed to be disolvate of the type Pd(NO3)2.2BESO. (author). 52 refs., 6 tabs., 6 figs

  6. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

    Mun, Ch

    2007-03-15

    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  7. New techniques for the treatment of laundry and other low-level liquid wastes

    PART I: The liquid wastes arising in BWR plants are generally classified under four types. Usually, three are reused as feed water after filtration, concentration and demineralization, while laundry liquid waste is discharged into the environment. Studies were made on laundry liquid waste treatment in order to establish a closed liquid waste system by using foamless detergent and evaporation treatment. It was demonstrated in pilot plant experiments that no foaming was caused during continuous concentration test runs from 500 to 200000 ppm. Then, the concentrated liquid was dried by LUWA wiped film vertical drier. The total volume reduction ratio through this concentration-drying system was calculated to be 1/2000. This treatment system may well be expected to have a potentiality of practical application in a few years to nuclear power plants in Japan. PART II: Although it is known that iron or iron compounds are effective in removing ruthenium, any satisfactory continuous treatment method of removing ruthenium present in minute quantities in the liquid waste has not yet been reported in the published literature. Based on the preliminary batch tests, column experiments using steel wool as the filter medium with activated surface layers produced by saturated steam were carried out mainly to develop a continuous method for industrial application. Some of the results of test runs made at a space velocity of 50m3.m-3.h-1, with 106Ru spiked laboratory demineralized water clearly showed that the DFs of more than 100 were obtained even after treatment of the volume of liquid 10000 times the bed volume. (author)

  8. Summary technical report on the electrochemical treatment of alkaline nuclear wastes

    This report summarizes the laboratory studies investigating the electrolytic treatment of alkaline solutions carried out under the direction of the Savannah River Technology Center from 1985-1992. Electrolytic treatment has been demonstrated at the laboratory scale to be feasible for the destruction of nitrate and nitrite and the removal of radioactive species such as 99Tc and 106Ru from Savannah River Site (SRS) decontaminated salt solution and other alkaline wastes. The reaction rate and current efficiency for the removal of these species are dependent on cell configuration, electrode material, nature of electrode surface, waste composition, current density, and temperature. Nitrogen, ammonia, and nitrous oxide have been identified as the nitrogen-containing reaction products from the electrochemical reduction of nitrate and nitrite under alkaline conditions. The reaction mechanism for the reduction is very complex. Voltammetric studies indicated that the electrode reactions involve surface phenomena and are not necessarily mass transfer controlled. In an undivided cell, results suggest an electrocatalytic role for oxygen via the generation of the superoxide anion. In general, more efficient reduction of nitrite and nitrate occurs at cathode materials with higher overpotentials for hydrogen evolution. Nitrate and nitrite destruction has also been demonstrated in engineering-scale flow reactors. In flow reactors, the nitrate/nitrite destruction efficiency is improved with an increase in the current density, temperature, and when the cell is operated in a divided cell configuration. Nafion reg-sign cation exchange membranes have exhibited good stability and consistent performance as separators in the divided-cell tests. The membranes were also shown to be unaffected by radiation at doses approximating four years of cell operation in treating decontaminated salt solution

  9. Off-gas treatment and characterization for a radioactive in situ vitrification test

    Effluents released to the off gas during the in situ vitrification (ISV) of a test site have been characterized. The site consisted of a 19 L waste package of soil containing 600 nCi/g transuranic and 30,000 nCi/g mixed fission products surrounded by uncontaminated soil. Radioactive isotopes present in the package were 241Am, /sup 238/239/Pu, 137Cs, 106Ru, 90Sr, and 60Co. The ISV process melted the waste package and surrounding soil and immobilized the radionuclides in place, producing a durable, 8.6 metric ton glass and crystalline monolith. The test successfully demonstrated that the process provides containment of radioactive material. No release to the environment was detected during processing or cooldown. Due to the high temperatures during processing, some gases were released into the off-gas hood that was placed over the test site. The hood was maintained at a light negative pressure to contain any volatile or entrained material during processing. Gases passed from the hood to an off-gas treatment system where they were treated using a venturi-ejector scrubber, a tandem nozzle gas cleaner scrubber followed by a condenser, heater, and two stages of HEPA filters. The off-gas treatment system is located in the semi-trailer to allow transport of the process to other potential test sites. Retention of all radionuclides by the vitrified zone was greater than 99%. Soil-to-off-gas decontamination factors (DFs) for transuranic elements averaged greater than 4000 and for fission products, DFs ranged from 130 for 137Cs to 3100 for 90Sr

  10. Environmental monitoring report on radiological status of the ground water beneath the Hanford site, January--December 1974

    Evaluation of β/sub t/, 3H and NO3/sup -/ concentrations measured in well water sampled near the surface of the unconfined ground water during 1974 shows that zones of contamination extend in an easterly to south-easterly direction from 200-E Area, as has been observed in the past. Tritium and NO3- ion concentrations above background were found along the Columbia River in the vicinity of the 100 Areas. A low ground-water mound showing uranium and nitrate ion concentrations above background continues to be observed in the 300 Area. Gross beta activity in the ground water beneath the 100 Areas was detectable only at the 100-N Area from effluent discharged to the 1301-N crib. A number of radionuclides, such as 106Ru, 60Co, 129I, 137Cs and 90Sr, occur at various locations and are detectable in very low concentrations in the unconfined ground water external to the 200 Areas. In most cases, concentrations are several orders of magnitude below applicable CG's and are so close to the detection limit as to cast doubt of their existence based on a single analysis. Only in and adjacent to the 200 Areas does any radionuclide that was analyzed in CY-1974 occur in concentrations greater than 10 percent of the applicable CG for uncontrolled water use. Nitrate ion concentrations greater than 100 percent of the Public Health Service (PHS) recommended drinking water standard of 45 mg/l continued to be observed in the ground water adjacent to the 200 Areas; in the vicinity of the 100-F and 300 Areas; and in the east-central part of the Hanford site (Wells 20-20, 26-15, and 32-22). Ground water from these zones is not consumed by human beings or other animals

  11. Data summary report for fission product release test VI-4

    This was the fourth in a series of high-temperature fission product release tests in a vertical test apparatus. The test specimen, a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium, had been irradiated to a burnup of 47 MWd/kg. In simulation of a severe accident in a light-water reactor, it was heated in hydrogen in a hot cell-mounted test apparatus to a maximum test temperature of 2400 K for a period of 20 min. The released fission products were collected on components designed to facilitate sampling and analysis. On-line radioactivity measurements and posttest inspection revealed that the fuel had partially collapsed at about the time the cladding melted. Based on fission product inventories measured in the fuel or calculated by ORIGEN2, analyses of test components showed total releases from the fuel of 85% for 85Kr, 106Ru, 3.9% for 125Sb, 96% for both 134Cs and 137Cs, and 13% for 154Eu. Large fractions of the released fission products (up to 96% of the 154Eu) were retained in the furnace. Small release fractions for several other fission products -- Rb, Br, Sr, Te, I, and Ba -- were detected also. In addition, very small amounts of fuel material -- uranium and plutonium -- were released. Total mass release from the furnace to the collection system, which included fission products, fuel material, and structural materials, was 0.40g, with 40% of this material being deposited as vapor and 60% of it being collected as aerosols. The results from this test were compared with previous tests in this series and with an in-pile test at similar conditions at Sandia National Laboratories. There was no indication that the mode of heating (fission heat vs radiant heat) significantly affected fission product release. 24 refs., 25 figs., 14 tabs

  12. Off-gas treatment and characterization for a radioactive in situ vitrification test

    Effluents released to the off gas during the in situ vitrification (ISV) of a test site have been characterized by Pacific Northwest Laboratory. The site consisted of a 19 L waste package of soil containing 600 nCi/g transuranic and 30,000 nCi/g mixed fission products surrounded by uncontaminated soil. Radioactive isotopes present in the package were 241Am, 238/239Pu, 137Cs, 106Ru, 90Sr, and 60Co. The ISV process melted the waste package and surrounding soil and immobilized the radionuclides in place, producing a durable, 8.6 metric ton glass and crystalline monolith. The test successfully demonstrated that the process provides containment of radioactive material. No release to the environment was detected during processing of cooldown. Due to the high temperature during processing, some gases were released into the off-gas hood that was over the test site. The hood was maintained at a slight negative pressure to contain any volatile or entrained material during processing. Gases passed from the hood to an off-gas treatment system where they were treated using a venturi-ejector scrubber, a tandem nozzle gas cleaner scrubber followed by a condenser, heater, and two stages of HEPA filters. The off-gas treatment system is located in the semi-trailer to allow transport of the process to other potential test sites. Retention of all radionuclides by the vitrified zone was greater than 99%. Soil-to-off-gas decontamination factors (DFs) for transuranic elements averaged greater than 4000 and for fission products, DFs ranged from 130 for 137Cs to 3100 for 90Sr. 7 references, 15 figures, 4 tables

  13. The state of the art on nuclides separation in high level liquid wastes by Truex process

    For the advancement of the back-end of nuclear fuel cycle, novel CMPO RUEX process was studied for separating minor actinides from fission products in high level liquid waste using real radioactive solutions from PUREX experiments, so as to support PNC's actinides recycling program using fast reactor. The present PUREX process was also studied to improve the separation of 237Np, 106Ru and 99Tc, the most interfering-natured nuclides in both PUREX and TRUEX processes, by utilizing electrochemistry-based salt-free methods which can eliminate the secondary radioactive waste. The state of the art of separation technologies are described by summarizing the extraction behaviors of nuclides in recent hot counter-current runs using CMPO RUEX process with mild salt-free stripping reagents. The degradation and regeneration characteristics of CMPO/TBP/n-dodecane mixture solvent were also simulated by semi-hot experiments. Several experiments to separate minor actinides and lanthanides from the TRUEX mixture product using aqueous amino-poly-carboxylate complexant, DTPA, resulted in reasonable MA/Ln separation profiles in multiple mixer-settler stages and allowed a unique separation flowsheet adaptable to the TRUEX process to be proposed. Application of electrochemistry to assist both solvent extraction processes, e.g., 'anodic oxidation' to destroy PUREX and TRUEX solvent waste in the presence of electron transfer mediator Age2+ or 'cathodic reduction' for electrolytic extraction of Pd2+, RuNO3+ and 99TcO4- from 3 M nitric acid medium is under study. (authors)

  14. The potentialities of the complexation ultrafiltration technique for the decontamination of fission product contaminated aqueous effluents; Potentialites de la complexation - ultrafiltration a la decontamination d`effluents radioactifs en produits de fission

    Thibert, V.

    1995-07-01

    Many nuclear researchers and industrial operators lay emphasis on improving the back end of the fuel cycle. A major problem concerns the liquid wastes generated by the reprocessing plant at La Hague, discharged into the sea after treatment in the Effluent Treatment Station (STE) (3), and which have become crucial matter. The activity of these wastes is well below the current legal limits, and is constantly decreasing these last years. To bring it close to zero, and ambitious goal, entails innovative new reprocessing techniques. We accordingly investigated the possibilities of complexation-ultrafiltration, a technique that uses water-soluble macromolecules to complex the target elements to be separated. We first achieved the strontium (II) separation with poly-acrylic and poly-sulfonic acids. The effects of pH and NaNO{sub 3} concentration influence on Sr (II) complexation were studied. The Sr (II) complexation and concentration phases, followed by cation de-complexation to recover the polymer, were also taken into account. This research, combined with a potentiometric study of the polymers, offered a close understanding of the chemical systems involved, and of the operating conditions and limits of complexation-ultrafiltration. The laboratory results were also validated on a tangential ultrafiltration pilot plant. We then used complexation-ultrafiltration to treat a real effluent generated bu La Hague`s STE 3 plant. This experiment demonstrated minimum 90 % decontamination of Sr (II) (with polyacrylate complexing agent), and also for {sup 134-137}Cs (with simple ultrafiltration). The use of two polyamides allowed partial decontamination of the effluent for {sup 60}Co and {sup 106}Ru. This work therefore offers a global approach to complexation-ultrafiltration, from laboratory to pilot scale, on real and simulated effluents. The future of this technique relies chiefly on the ability to solve the problem of polymer recovery. (Abstract Truncated)

  15. Laboratory development of methods for centralized treatment of liquid low-level waste at Oak Ridge National Laboratory

    Improved centralized treatment methods are needed in the management of liquid low-level waste (LLLW) at Oak Ridge National Laboratory (ORNL). LLLW, which usually contains radioactive contaminants at concentrations up to millicurie-per-liter levels, has accumulated in underground storage tanks for over 10 years and has reached a volume of over 350,000 gal. These wastes have been collected since 1984 and are a complex mixture of wastes from past nuclear energy research activities. The waste is a highly alkaline 4-5 M NaNO3 solution with smaller amounts of other salts. This type of waste will continue to be generated as a consequence of future ORNL research programs. Future LLLW (referred to as newly generated LLLW or NGLLLW) is expected to a highly alkaline solution of sodium carbonate and sodium hydroxide with a smaller concentration of sodium nitrate. New treatment facilities are needed to improve the manner in which these wastes are managed. These facilities must be capable of separating and reducing the volume of radioactive contaminants to small stable waste forms. Treated liquids must meet criteria for either discharge to the environment or solidification for onsite disposal. Laboratory testing was performed using simulated waste solutions prepared using the available characterization information as a basis. Testing was conducted to evaluate various methods for selective removal of the major contaminants. The major contaminants requiring removal from Melton Valley Storage Tank liquids are 90Sr and 137Cs. Principal contaminants in NGLLLW are 9OSr, 137Cs, and 106Ru. Strontium removal testing began with literature studies and scoping tests with several ion-exchange materials and sorbents

  16. Relation between physicomechanical properties and diffusion phenomena in composite materials

    One of the procedures for storing low and medium activity nuclear waste consists of coating the contaminated material in a thermosetting resin. The drums thus constitued are stored in concreted underground trenches, then covered with cement, bitumen or clayey soil. Although the risk of water circulation is low, this element represents on the one hand the major cause of natural deterioration of the polymer, and on the other hand the most likely vehicle for conveying the radioactive ions confined in the drums. It is for this reason that the study of the behaviour of polyester or epoxide-based macromolecular materials with regard to water constitutes the first stage of this work. The second part of the thesis is directed towards the study of compound materials. Indeed, the charges are represented in the first case by the nuclear waste itself; in the second case, they are introduced into the polymer beforehand, on the one hand to reduce costs, and on the other hand to give the mixture suitable mechanical and rheological properties. In this study, three types of mineral charge are added in an epoxide resin: glass balls surface-treated or not, and sand. Various techniques are implemented in order to assess and characterize the interfacial adhesion, in the different systems. The strongest polymer-charge bonds are sought in order to resist natural deterioration. Finally, the object of the confinement process, is to avoid dispersion of low and medium activity substances (137Cs, 90Sr, 60Co, 106Ru..) in the environment. The final stage of this work therefore consists in assessing the barrier qualities of pure or charged polymers with regard to radioactive ion diffusion. We will show in particular that the use of fine resin membranes enables the diffusion coefficient of the 137Cs to be calculated

  17. Biokinetics and dose assessment of radionuclides in juveniles

    In the aftermath of the Chernobyl accident, it becomes evident that dose coefficients for members of the public are necessary. International Commission on Radiological Protection (ICRP) established a task group of Committee 2 charged with the assessment of dose coefficients as a function of an individual's age. However, little data is available on the biokinetics of radionuclides in juvenile and there is a need to develop age-dependent biokinetic models, such as for the gastrointestinal tract. The present paper reviewed an outline on characteristics of biokinetics of radionuclides in juvenile animals focusing on the previous experimental data. The following radionuclides are discussed: 54Mn, 60Co, 65Zn, 75Se, 106Ru, 110mAg, 115mCd, 125Sb, 137Cs, 141Ce, 203Hg and 3H. Generally, intestinal absorption and whole-body retention of radionuclides in juveniles were higher than that of adult. In the case of sucklings, it is very important to study how radionuclides are transferred through the placenta and milk. The transfer rate of radionuclides through the placenta and milk is dependent on the period of gestation at the time of dosing. The IDES(Internal Dose Estimation System) which is based on the ICRP model was used for dose calculation. We modified the IDES using the biokinetic data which was gained animal experiment. The IDES is flexible because the absorbed dose can be calculated by substituting arbitrary physical and physiological parameters and also substituting ingested dose coefficients not only for the ICRP Reference Man, but also for Japanese of 1 year old, 5 years old, 10 years old, 15 years old and the adult, respectively. (author)

  18. Experimental Fission Gas Release Determination at High Burnup by Means of Gamma Measurements on Fuel Rods in OL2

    This article presents the results from the gamma measurements performed on a selection of fuel rods from two SVEA-96 Optima fuel assemblies in Olkiluoto unit 2 (OL2) during February 2008. The measurements were funded by Teollisuuden Voima Oyj (TVO) and carried out by Westinghouse Electric Sweden AB (WSE). The goal of the measurements was to obtain plant specific fission gas release data from OL2 which will later be used to support TVO's burnup increase. The measurements were performed by means of detecting and recording information on gamma rays emanating from radioactive fission products within the fuel rods. A fuel assembly under operation in the reactor will be subject to fission gas release, meaning that gaseous fission products in the fuel matrix will leak out into the fuel rod free volumes, including the upper fuel rod plenum. The magnitude of fission gas release is closely related to the operation conditions, such as burnup and the pellet power/temperature history. Fission gas which is released into the plenum leads to an increased pressure within the fuel rod. Since the gas plenum is a volume free of fuel pellets, radioactive gases present here are therefore relatively easy to measure. A suitable gaseous radioactive fission product is 85Kr which has a gamma energy of 514 keV, and a half life of 10.8 years. Measuring the amount of 85Kr can then be used to quantify the amount of fission gases release during operation. Other sources of gamma rays in the plenum with similar energies are the β+-emitting 58Co and the 106Ru. The half lives of 58Co and 106Ru are 71 days and 372 days, respectively, with corresponding gamma ray energies of 511 keV and 512 keV, respectively. In total, three different gamma rays with similar energies must be resolved by the detector system. In order to perform the measurements, 58Co must have decayed to an extent that allows the 514 keV line to be resolved from the 511 keV positron annihilation gamma ray. The 106Ru is, on the other

  19. Deposition of Chernobyl-derived transuranium nuclides and short- lived radon-222 progeny in Finland

    In this study the atmospheric deposition of radionuclides was investigated from two different viewpoints. The Chernobyl nuclear power plant accident in April 1986 caused a widely spread plume of radionuclides, including transuranium elements. The regional deposition of these elements in Finland was assessed based on lichen and peat samples. Unlike the deposition of transuranium elements from the weapons tests in the 1950's and 1960's, the deposition from the Chernobyl accident was very unevenly distributed in Finland. Also, the Chernobyl-derived deposition of 299,240Pu even in the most contaminated regions in Finland was still only some 10 per cent of the global fallout from weapons tests. On the other hand, the measured activity concentrations of 241Pu in the upper parts of lichen samples are comparable to those found in samples comparable during the heaviest weapons-test fallout in the early 1960's. The observed average 241Pu/239,240Pu activity ratio in the upper parts of lichen, 95, can be expected to lead at its maximum in the year 2059 to 241Am/239,240Pu activity ratio of 2.8 in the Chernobyl-derived deposition, exclusive of the 241Am present in the original deposition in 1986. The deposition pattern of transuranium elements observed in this work resembles that of refractory gamma-emitting nuclides such as 95Zr and 141Ce. The sampling area of this investigation does not cover the northern part of Finland. However, the fallout pattern of 95Zr would suggest that the deposition of transuranium nuclides north of the 65th latitude was very low. Biological half-lives of 730 d and 320 d for Pu and Am, respectively, were obtained in lichen in this study. The second part of this work is concerned with the factors affecting the wet deposition efficiency of the natural short-lived radon-222 progeny. This was studied with two methods: using recordings of external gamma radiation in central Finland, and using an automatic precipitation gamma analyser in northern Finland

  20. 放射性核素对生态环境的影响%NUCLEAR ACCIDENT IMPACT ON THE ECOLOGICAL ENVIRONMENT

    史建君

    2011-01-01

    This article reviewed the eeo-environmental behavior of radionuclides released into the environment by nuclear explosion and nuclear accidents, especially of several key radionuclides with biological significance, including 137 Cs,95Zr, 90Sr, 131I,3H and 14C, in order to correctly understand the case of nuclear accidents and its pollution, maintain the social stable, and provide suitable measures for environmental protection and safety.%本文概述了核爆炸以及发生重大核事故时进入生态环境的放射性核素,特别是具生物学意义的几种重点核素的生态环境行为,主要包括137Cs、95Zr、90Sr、131I、3H和14C,为正确认识核事故情况下的污染状况,维持社会稳定,对制定应急计划,采取积极可行的应对措施提供有益的借鉴.

  1. Study of non-1/ ν reaction nuclides using k0 - Neutron Activation Analysis at the Malaysian Nuclear Agency Research Reactor

    The modified spectral index r(α); the Westcott gLu(Tn) factor and absolute neutron temperature Tn were determined for the handling of non-1/ ν (n, γ) reaction based on the Westcott formalism using k0-neutron activation analysis (k0-NAA) method at the Malaysian Nuclear Agency (MNA) research reactor. The r(α) was determined by the bare bi-isotopic monitor method using measurement of radionuclides of 97Zr and 95Zr. The 176Lu as non-1/ ν and 197Au as 1/ ν monitors were utilized for determination of gLu(Tn). The r(α) and gLu(Tn) values ranged from 0.0715 to 0.1417 with a RSD of 15.24 % and from 1.7832 to 2.0149 with a RSD of 3.58 %, respectively. The accuracy of the method was evaluated based on the calculated absolute neutron temperature (Tn) value. The calculated average value of Tn was 40.56 ± 9.32 degree Celsius while the value reported by MNA was 40 degree Celsius, which represents an acceptable level of consistency. (author)

  2. The use of algae in monitoring discharges of radionuclides. Experiences from the 1992 and 1993 monitoring programmes at the Swedish nuclear power plants

    All four Swedish nuclear power plants (Forsmark, Oskarshamn, Barsebaeck and Ringhals) use brackish water as coolant (Baltic Sea and Swedish west coast). Radionuclides are discharged together with the cooling water. The gamma spectra of monthly algal samples harvested in 1992 and 1993 close to the discharge points of these power plants were determined within the routine monitoring programmes. The main radionuclides detected in the algal samples were 54Mn, 58Co, 60Co and 137Cs. Most 137Cs in the samples from the northern Baltic Sea (Forsmark) still originated from the 1986 Chernobyl accident. Other radionuclides, notably 51Cr, 65Zn, 95Zr, 95Nb, 110mAg, 124b, 125Sb and 134Cs, were regularly detected at s of the sites. Transfer factors from discharge to algae were generally in the order of 0.3-3 Bq kg-1 per MBq discharge. For the major discharged radionuclides, significant linear relationships were in most cases found between discharges and concentrations in algal samples. Differences in transfer factors and regression coefficients were explained by location of the sampling sites and type of radionuclide. It is concluded that algal samples provide useful complements to water and sediment samples in the monitoring programmes since radionuclide concentrations are much higher in algal samples and proportional to the discharges. 21 refs, figs

  3. Countercurrent flowsheet testing of the TRUEX process with ICPP calcine

    Law, J.D.; Herbst, R.S.; Brewer, K.N.; Todd, T.A.

    1998-07-01

    Calcine was generated at the Idaho Chemical Processing Plant over several decades as a method of solidifying numerous raffinates and wastes from spent nuclear fuel reprocessing for convenient interim storage. Unfortunately, the bulk of the calcine is inert, with radionuclides comprising less than 1 weight percent of the total calcine mass. The bulk of the calcine currently stored at the ICPP was produced from wastes generated during reprocessing of zirconium clad fuels. Consequently, this material contains varying, but large quantities of zirconium oxide. Currently, separations options are being considered for acidic solutions of dissolved ICPP calcines to minimize high level waste volumes and economic penalties perceived for final disposal of these wastes. The actinide separation process being emphasized for the dissolved calcine solutions is the TRUEX process. Substantial problems have been encountered during TRUEX flowsheet development efforts for dissolved zirconium calcine simulant due to the high concentrations and subsequent extraction of zirconium from the feed. Alteration of the calcine dissolution parameters has resulted in the development of a successful TRUEX/Zr calcine baseline flowsheet. This flowsheet has been tested using 22 stages of a 2.0 centimeter diameter centrifugal contactor pilot plant using simulated dissolved Zr calcine solution. With this flowsheet, a removal efficiency of > 96% was obtained for {sup 241}Am (analytical detection limits were reached). Less than 0.25% of the {sup 95}Zr exited with the high-level waste strip product.

  4. Preparation of nuclear pure zirconium oxide from zircon

    Nuclear fuel used in the commercial nuclear reactors is cladded to confine the radioactivity. Zirconium based alloys standout as cladding materials because of their high mechanical strengh at high temperatures and pressures combined with good corrosion resistence and a low absorption cross section for thermal neutrons. However, a separation procedure to reduce the Hafnium content which occurs along with Zirconium and possesses a high neutron absorption cross section is needed. The preparation of nuclear pure Zirconium from Zircon is presented. The mineral was opened by alkali fusion at 4500C and later transformed into Zirconyl Nitrate via oxychloride and basic carbonate and purified by solvent extraction with TBP-HNO3. The solvent extraction process was developed using 95Zr and 181Hf tracers and studing process variables like acidity, nitrate and metal concentration. The extraction of Zirconium with 60% TBP in kerosene equilibrated with 5 M HNO3 increases with increasing acidity and nitrate concentration. The dependence of coefficient of distribution with acidity was of the power = 1.5 and with concentration of nitrate was of 3 sup(rd) power under the experimental conditions. The extraction of Zirconium and Hafnium reduces with increasing loading of the solvent but the separation factor remained approximately constant. (Author)

  5. Transfer coefficients of selected radionuclides to animal products. I. Comparison of milk and meat from dairy cows and goats

    The diet-milk transfer coefficient, Fm (Bq L-1 output in milk divided by Bq d-1 intake to the animal) was studied for eight radionuclides that previously had been given little attention. The Fm values for cows and goats, respectively, were: 2.3 x 10(-5) and 1.5 x 10(-4) for /sup 99m/Tc, 1.4 x 10(-4) and 8.5 x 10(-4) for /sup 95m/Tc, 1.1 x 10(-2) for 99Tc (goats only); 1.7 x 10(-3) and 9 x 10(-3) for 99Mo; 4.8 x 10(-4) and 4.4 x 10(-3) for /sup 123m/Te; 4.8 x 10(-4) and 4.6 x 10(-3) for 133Ba; 5.5 x 10(-7) and 5.5 x 10(-6) for 95Zr; and 4.1 x 10(-7) and 6.4 x 10(-6) for 95Nb. The goat/cow transfer coefficient ratios for milk were approximately 10, but the goat/cow ratios for meat varied by three orders of magnitude

  6. New activation cross section data on longer lived radio-nuclei produced in proton induced nuclear reaction on zirconium

    The excitation functions of 96Nb, 95mNb, 95gNb, 92mNb, 91mNb, 90Nb, 95Zr, 89Zr, 88Zr, 86Zr, 88Y, 87mY, 87gY, 86Y were measured up to 70 MeV proton energy by using the stacked foil technique and the activation method. The new data were compared with the critically analyzed experimental data in the literature and with the TALYS based model results in TENDL-2013 library. The possible role of the investigated reactions in the production of medically relevant 90Nb, 95mNb, 89Zr, and 88Y radionuclides is discussed. - Highlights: • Proton induced reactions on natural zirconium up to 65 MeV. • Stacked foil irradiation technique coupled with gamma-spectrometry. • Comparison of experimental data with the nuclear reaction model results in the TENDL-2013 library. • Calculation and comparison of thick target integral yields. • Comparison of the production routes of 90Nb, 95mNb, 89Zr and 88Y medically relevant radioisotopes

  7. Varieties of fuel particles in fallout of the Chernobyl NPP exclusion zone

    Soil in the Chernobyl NPP 30-km zone is known to be contaminated by fine particles of dispersed irradiated fuel. Experimental data obtained by the authors shortly after the accident indicated a significant excess of 95Zr in the particles compared with its calculated amount. This suggested that particles of the U fuel also contain construction material from the core, activated Zr. The authors have isolated particles of radioactive material from the heavy soil fraction in order experimentally to confirm this. Their dimensions varied from 85 to 750 μm; their mass, from 10 to 700 μg. Such characteristics as the microstructure, elemental composition, and the concentration of 144Ce and other radionuclides obtained from electron microscope measurements, microprobe analysis, and γ- and α-spectrometry were used to classify the particles. It was found that two types of highly active particles are characteristic of the Chernobyl NPP exclusion zone (up to 4-5 km) and the western track of the fallout for a distance of 20 km from the destroyed fourth block. The first type are genuine particles of irradiated fuel. The second type are aggregates consisting of U and Zr oxides. The second type predominates in fallout of the exclusion zone and in the western track

  8. The Chernobyl accident - a meteorological analysis of how radionuclides reached and were deposited in Sweden

    The atmospheric transport to Sweden and the deposition of radioactive material following the Chernobyl accident have been described on the basis of radiological and meteorological data and theoretical calculations of dispersion. The radioactive cloud created by the explosion at 01.23 local time on April 26, 1986, was transported northwest and north over the Baltic Sea, with the first radionuclides probably reaching southeast Sweden early on April 27. In Svealand (eastern central Sweden), high concentrations of radionuclides began to appear on the evening of the same day. Sweden was affected by dry deposition, including fairly large hot particles and also by wet deposition. Wet deposition occurred via precipitation over eastern Sweden on April 28 and over parts of northern Sweden until April 30. The deposition of radionuclides over the country was mapped in detail, by extensive measurements of gamma radiation using aircraft and by in situ spectroscopy on the ground. Deposition of cesium mainly occurred through wet deposition. In the case of other nuclides, for example, 95Zr and 230Np, a considerable part occurred as dry deposition. An integration of the total cesium deposition shows that as much as about 10 percent of the total quantity of cesium released from Chernobyl may have been deposited over Sweden

  9. α-Zr self-diffusion anisotropy

    Self-diffusion coefficients (D) have been measured in nominally pure (NP) α-Zr single crystals (∼ 50 ppma Fe) in the range 867-1107 K, in directions either parallel (Dpa) or perpendicular (Dpe) to the c-axis. Measurements were also made on high-purity (HP) α-Zr single crystals (95Zr) counting. Sectioning was done with a sputtering device, or a microtome (some NP experiments at 1107 K). D values for NP Zr are about an order of magnitude higher than the corresponding values for HP Zr. Diffusion anisotropy is complicated. The sputter-sectioned NP Zr specimens show increasing anisotropy ratios (AR = Dpa/Dpe), from 1.0 to 3.2, with decreasing temperatures, whereas AR = 0.53 for both the microtome-sectioned NP and sputter-sectioned HP Zr: the low AR value is consistent with expectations based on intrinsic self-diffusion in hcp metals with c/a < 1.633. (author). 12 refs., 1 tab., 3 figs

  10. Whole-body gamma-spectroscopic assessment of environmental radionuclides in recapturable wild birds

    Free-ranging recapturable populations of wild birds trapped at a nuclear power station and at a number of control sites were measured for gamma-emitting radionuclides. Body burdens were ascertained with a specially designed whole-body counter. After being banded and counted, birds were returned to the site of capture for release and subsequent recapture. Birds showing unusually high radionuclide levels were held for the determination of the biological half-life of the detected nuclide. Bobwhite and blue jays showed 137Cs body burdens which exhibited significant variations between sites separated by as little as 9 km. A marked temporal decrease in the 137Cs concentrations was recently seen in a number of species only at the reactor trapping sites. The biological half-life of 137Cs in the Blue Jay is approximately 6.7 days. Transient occurrences of 131I and 95Zr--95Nb are reported and summary radionuclide data for 15 passerine species are presented. To date there is no evidence to link the routine operation of the nuclear power plant to the observed avian radionuclide concentration data

  11. Radioactivity Levels in Kola Bay

    Sediment samples were collected in May 1995 from 16 locations in Kola Bay, North-west Russia, during an expedition starting from Murmansk and ending at Kildin Island in the Barents Sea. The purpose was to study the contamination level in an area with several potential sources of civilian and military radioactive pollution. 137Cs concentrations in the sediments, algae and benthic samples were low, but small particles containing 137Cs were separated from the sediment samples. All the sediments between the nuclear icebreaker base Atomflot and the open Barents Sea contained 60Co. Traces of 125Sb, 134Cs, 95Zr, 154Eu and 152Eu were also detected in some of the samples. Plutonium levels were low, but the increased 238Pu/239,240Pu ratio at Atomflot indicated a fresh release from the facility or from the waste storage vessels, Lepse and Imandra, lying in front of it. An increased 238Pu/239,240Pu ratio was also found in sediment collected in the outlet of Kola Bay in the Barents Sea. (author)

  12. The impact of updated Zr neutron-capture cross sections and new asymptotic giant branch models on our understanding of the s process and the origin of stardust

    Lugaro, Maria; Karakas, Amanda I; Milazzo, Paolo M; Kaeppeler, Franz; Davis, Andrew M; Savina, Michael R

    2013-01-01

    We present model predictions for the Zr isotopic ratios produced by slow neutron captures in C-rich asymptotic giant branch (AGB) stars of masses 1.25 to 4 Msun and metallicities Z=0.01 to 0.03, and compare them to data from single meteoritic stardust silicon carbide (SiC) and high-density graphite grains that condensed in the outflows of these stars. We compare predictions produced using the Zr neutron-capture cross section from Bao et al. (2000) and from n_TOF experiments at CERN, and present a new evaluation for the neutron-capture cross section of the unstable isotope 95Zr, the branching point leading to the production of 96Zr. The new cross sections generally presents an improved match with the observational data, except for the 92Zr/94Zr ratios, which are on average still substantially higher than predicted. The 96Zr/94Zr ratios can be explained using our range of initial stellar masses, with the most 96Zr-depleted grains originating from AGB stars of masses 1.8 - 3 Msun, and the others from either lowe...

  13. Determination of uranium fission products interference factors in neutron activation analysis

    Neutron activation analysis is a method used in the determination of several elements in different kinds of matrices. However, when the sample contains high U levels the problem of 235U fission interference occurs. A way to solve this problem is to perform the correction using the interference factor due to U fission for the radionuclides used on elemental analysis. In this study was determined the interference factor due to U fission for the radioisotopes 141Ce, 143Ce, 140La, 99Mo, 147Nd, 153Sm and 95Zr in the research nuclear reactor IEA-R1 on IPEN-CNEN/SP. These interference factors were determined experimentally, by irradiation of synthetic standards for 8 hours in a selected position in the reactor, and theoretically, determining the epithermal to neutron fluxes ratio in the same position where synthetic standards were irradiated and using reported nuclear parameters on the literature. The obtained interference factors were compared with values reported by other works. To evaluate the reliability of these factors they were applied in the analysis of studied elements in the certified reference materials NIST 8704 Buffalo River Sediment, IRMM BCR- 667 Estuarine Sediment e IAEA-SL-1 Lake Sediment. (author)

  14. Foil activity measurements for testing transport calculations in the Budapest Research Reactor

    The upgraded VVR-SM type (Russian design) Budapest Research Reactor serves both research and practical applications. As a by-product of the experimental methods used in the field of the neutron activation analysis a unique opportunity arose for benchmarking the neutron physical algorithms against measurements. As the original aim of the measurements was the determination of the concentrations and the necessary neutron flux characteristics, the measured primary data had to be reevaluated to verity the neutron physical calculations. The reaction rates of the following measured reactions were selected for the comparison: 94Zr(n,γ)95Zr, 96Zr(n,γ)97Zr/97mNb, 58Ni(n,p)58Co, 176Lu(n,γ)177Lu and 197Au(n,γ)198Au. For the sake of comparison, the multigroup cross section library of the MULTICELL code had been supplemented with the data of the above reactions by using the NJOY code. As the reaction rates are measured at the same positions (practically without shielding effects), the measured and calculated reaction rate ratios were compared on the level of the multigroup MULTICELL calculations. The accuracy of the MULTICELL code for the research reactor has been tested by comparative MCNP calculations. (author)

  15. Determination of uranium fission products interference factors in neutron activation analysis; Determinacao de fatores de interferencia de produtos de fissao de uranio na analise por ativacao neutronica

    Ribeiro Junior, Ibere Souza

    2014-09-01

    Neutron activation analysis is a method used in the determination of several elements in different kinds of matrices. However, when the sample contains high U levels the problem of {sup 235}U fission interference occurs. A way to solve this problem is to perform the correction using the interference factor due to U fission for the radionuclides used on elemental analysis. In this study was determined the interference factor due to U fission for the radioisotopes {sup 141}Ce, {sup 143}Ce, {sup 140}La, {sup 99}Mo, {sup 147}Nd, {sup 153}Sm and {sup 95}Zr in the research nuclear reactor IEA-R1 on IPEN-CNEN/SP. These interference factors were determined experimentally, by irradiation of synthetic standards for 8 hours in a selected position in the reactor, and theoretically, determining the epithermal to neutron fluxes ratio in the same position where synthetic standards were irradiated and using reported nuclear parameters on the literature. The obtained interference factors were compared with values reported by other works. To evaluate the reliability of these factors they were applied in the analysis of studied elements in the certified reference materials NIST 8704 Buffalo River Sediment, IRMM BCR- 667 Estuarine Sediment e IAEA-SL-1 Lake Sediment. (author)

  16. Evaluation of fission product yields from fission spectrum n+239Pu using a meta analysis of benchmark data

    Chadwick, Mark B.

    2009-10-01

    Los Alamos conducted a dual fission-chamber experiment in the 1970s in the Bigten critical assembly to determine fission product data in a fast (fission neutron spectrum) environment, and this defined the Laboratory's fission basis today. We describe how the data from this experiment are consistent with other benchmark fission product yield measurements for 95,97Zr, 140Ba, 143,144Ce, 137Cs from the NIST-led ILRR fission chamber experiments, and from Maeck's mass-spectrometry data. We perform a new evaluation of the fission product yields that is planned for ENDF/B-VII.1. Because the measurement database for some of the FPs is small—especially for 147Nd and 99Mo—we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data. The %-relative changes compared to ENDF/B-VI are small for some FPs (less than 1% for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (3%) and 147Nd (5%). We suggest an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average energies.

  17. Hydrogen migration in pure polycrystalline zirconium

    Full text: Pure (99.95 %) Zr was studied in a forced pendulum at three different conditions: 1) as received; 2) after hydriding the sample inside the pendulum, by exposing it to a hydrogen gas pressure of 60 kPa at 290 K during 1 h; and 3) after additional hydriding at 173 K during 8 hours. At the end of internal friction measurements, the remaining hydrogen content in each sample was determined by a destructive test inside a commercial equipment. The determined H contents were 8.6 wppm H in case (1), and 36 wppm H in case (3). Several peaks are detected from 50 μm grain size polycrystalline samples (1 mm diameter, 20 mm length). The pendulum measurements were taken in the (5-100) K temperature range at (1-25) Hz. The samples were surrounded by a 0.2 bar He-gas atmosphere ensuring good thermal transfer. For all cases the results over the spectra presented similar peaks near 40 K, which are discussed in connection with the H tunneling effect. Atomistic simulations of the embedded atom type were performed on the Zr-H system, and the calculated hydrogen migration energies are in good agreement with internal friction interpretations. (author)

  18. External radiation exposure of residents living close to the Mayak facility: main sources, dose estimates, and comparison with earlier assessments.

    Mokrov, Yury G

    2004-07-01

    In 1951 and 1952 specialists from the Mayak production association investigated the radiological situation in the area of the Metlinski reservoir that was located 5-7 km from the site of liquid radioactive waste (LRW) discharge. Based on their measurements of both the specific radioactivity in the water and the dose-rate above the water surface, the gamma-field above the water surface in 1951 was demonstrated to be mainly due to (95)Zr+(95)Nb. The dose-rate at the shore of the reservoir was calculated for the period 1949-1951. In November and December 1951, the gamma-field at the shore was mainly due to (140)Ba+(140)La. For the period 1949-1951, the external exposure of the Metlino population due to the decay of these radionuclides was about 200 R (2 Sv), most of the dose having been produced in 1951. The contribution of (137)Cs to external doses did at that time probably not exceed a fraction of several percent. This finding is in contradiction to the assumptions made in the most recent TRDS-2000 system that was developed to reconstruct the doses to the residents of the Techa river. The results presented here demonstrate that the reconstruction of external doses received by the Metlino population as well as by the Techa river residents can be improved for the most critical period between 1949 and 1954. PMID:15221313

  19. Peculiarities of radionuclide migration in main links of water reservoir

    The radionuclide uptake in fish tissues obeys general regularities. Species, fish age, diet character, ratio of quantities of water and food, temperature, a season, degree of water mineralization as well as chemical properties of an element to which the radionuclide belongs, in particular, its permeability through semipenetrable partitions influence the level of uptake. Radionuclides are taken up into the fish bodies from water by the mills, skin tissues, as well as through food. Experimental studies following a single contamination of water reservoirs with nuclear fission products have shown that the time for the maximum uptake of different nuclides in the fish tissues is rather different and varies from 3 to 35 days. The maximum uptake of 90Sr, whose concentration decreases slowly in water because of its weak sorption by soil, is attained later than that of 91Y and 95Zr, whose concentration in water decreases quickly due to their sorption by suspended-in-water soil particles and due to sedimentation of the latter

  20. Wear measurement using radioactive tracer technique based on proton, deuteron and α-particle induced nuclear reactions on molybdenum

    Highlights: ► Proton, deuteron, 3He and α-particle activation of Mo. ► TLA (thin layer activation). ► Wear measurement. ► Integral production yields. ► Wear curves (specific activity versus penetration depth). - Abstract: Excitation functions of light ion induced nuclear reactions on natural molybdenum have been studied in the frame of a systematic investigation of charged particle induced nuclear reactions on metals for various applications. Excitation functions of 93,94g,94m,95g,95m,96,99mTc, 90,93m,99Mo, 90,91m,92m,95m,95g,96Nb and 88,89Zr were measured up to 50 MeV deuteron energy Tárkányi et al., 2012 [1], 93m,93g,94m,94g,95m,95g,96g,99mTc, 90,93m,99Mo, 90,92m,95m,95g,96Nb and 88,89Zr were measured up to 40 MeV proton energy Tárkányi et al., 2012 [2] and 93m,93g,94m,94g,95m,95g,96g,99mTc, 93m,99Mo, 90Nb, 94,95,97,103Ru and 88Zr were measured up to 40 MeV alpha energy Ditrói et al., 2012 [3] by using the stacked foil technique and activation method. The results for 3He induced reactions on natural Mo were taken from the literature Comparetto and Qaim, 1980 [4]. According to their half-lives, from the above listed radionuclides the 95m,96Tc, 91m,92m,95m,95gNb, 99Mo, 103,97Ru and 88Zr are suitable candidates for wear measurement by using thin layer activation (TLA) method. The goal of this work was to determine the necessary nuclear data for TLA of the above radionuclides and to prove their applicability for wear measurements.

  1. Fission Yields of Some Isotopes in the Fission of Th232 by Reactor Neutrons

    The fission yields of the longer-lived isotopes produced in the fission of Th232 are not very well known; existing data show rather large discrepancies and/or uncertainties. Since we feel that at least some of these discrepancies arise from difficulties in measuring the absolute activities of the fission products, we measured the fission yield of 10 selected isotopes whose decay schemes are well understood. The thorium foils were irradiated in a position at the edge of the core of the SAPHIR swimming pool reactor. Following irradiation, the thorium was dissolved after addition of appropriate carriers. The fission products of interest were determined by conventional radiochemical methods that had to be modified slightly to ensure good decontamination from the abundantly formed Pa233 . The chemical yields were determined by gravimetric methods. Counting was done preferentially on a γ-spectrometer that had been calibrated at 11 different energies by standards either obtained from the IAEA or prepared by 4πβ-counting. In the case of Sr90, Ru106 and Ce144 a β-proportional counter was used that had been calibrated for these isotopes. In addition to the sought elements, Mo99 was isolated from each foil to serve as an internal monitor for the number of fissions taking place. The experiment thus gave the ratio of the yield of the sought element to the yield of Mo99. This ratio ''R'' was obtained for Sr90, Ru103, Ru106, Ag111, Pd112, I131, Cs137, Ba140, Ba141, Ce141 and Ce144, Results indicate the existence of a third peak in the yield mass curve in the region of symmetric fission. Yields of fission products relative to the Mo99 yields are given, and the absolute yields calculated by assuming y Mo99 = 2.78%. This number was derived from the work of Iyer et al., and was obtained by normalizing the area under the yield mass curve to 200%. (author)

  2. R[ionuclide transport after the Chernobyl reactor accident and derivation of r[ioecological parameters

    branches, mykorrhiza mushrooms, forest honey and roe and red deer meat is remarkable. The single time sequences show a fairly closed picture of the transport of the r[ionuclides in the environment which could be measured in Aachen with normal expenditure. These were in particular: 103Ru, 131I, 132Te, 134Cs and 137Cs. Most of the measured values were summarized in voluminous tables and diagrams. (orig.)

  3. Movements and distributions of radionuclide released from Chernobyl in the biosphere of Biwa lake

    On 26 April 1986 an explosion occurred at the Chernobyl nuclear plants in USSR which led to a global dispersion of radioactivity. We prepared the environmental samples collected from Biwa-lake which is the largest lake and located in the center of Japan. Radio-activities of the samples were measured in order to estimate the radioactive contamination in Japan. 1) The radio-nuclides in the water residues collected after one month of the accident were I-131, Ru-103, Ru-106 (Rh-106), Cs-134 and Cs-137. However, after half a year only Cs-137 was detected. The concentration had been reduced to less than one-tenth of the initial measured value. 2) Radio-nuclides detected in the bottom mud were Cs-137 and natural radioactive nuclides, i.e. K-40, thorium decay products (Tl-208, Ac-228) and uranium decay products (Pb-214, Bi-214). Cs-137 depth distribution in the bottom mud showed a maximum at 10-20 cm under the bottom surface, especially high in the clay-rich mud. Most of the Cs-137 concentration are estimated to be mainly due to the fallout before the Chernobyl accident, because a little of Cs-134 was detected in the only surface mud. It was reported by L. Devell et al. that the Cs-134/Cs-137 ratio originated from the Chernobyl accident, was 1/2 at the accident. 3) In the lake organisms Cs-134, Cs-137, Zr-95 and a trace of Ag-110m were detected. Among the all samples, Cs-137 in the flesh of black basses showed the highest value, 2.0 Bq/kg. Existence of Cs-134 shows that a part of the cesium originated from the Chernobyl accident. 4) The concentrations of Cs-137 in the flesh of black basses and carps were 2.0 Bg/kg and 0.37 Bq/kg after a half year of the accident, respectively. The Cs-137 concentration of black bass decreased to 1.45 Bq/kg after one year, 1.1 Bq/kg after 2 years (June 1988), and 0.50 Bq/kg after 5 years of the accident. The apparent effective half life of Cs-137 in the flesh of black basses was estimated to be about one year, although the radioactivity of

  4. Assessment of doses to biota in the river system

    Doses to aquatic biota in the hydrological system Techa - Ob are estimated.The following water bodies with different levels of radioactive contamination are considered: industrial reservoirs, Techa, Iset, Tobol and Irtysh Rivers. Doses to biota are calculated using the observed data on the content of radionuclides in various environmental components, with consideration for geometric characteristics of the organisms and the exposure sources. The following groups of the river biota are considered: aquatic plants, mollusks and fish. Simplified geometric models (ellipsoids) are used in the internal dose calculations for fish and mollusks. Aquatic plants are approximated either with spheres or with a layer of finite depth. For the external doses assessment the water was considered as an infinite source with the uniform distribution of radionuclides. Sediments were represented as a source with the uniformly distributed activity. Concentration factor of scattered radiation was taken into account for gamma emitters. Sources and levels of radioactive contamination of the Techa - Ob system are analyzed. Data on the activity concentration of radionuclides in water, bottom sediments and aquatic biota are used for the dose assessment. Assessment of doses to biota in the Techa -Ob river system in the period from 1949 to the present time are performed.The highest doses (over 0.01 Gy/day) were received by aquatic organisms in the upper reaches of the Techa River in the period of maximum discharges of radionuclides (1950-1951). In that period, a major contribution to the dose to aquatic organisms was due to the incorporated radionuclides: 89 Sr, 90 Sr, 106 Ru,137 Cs, 144 Ce and others. During 1950-1951, the doses to aquatic organisms were estimated, on average, at 0.003-0.1 Gy/day. After the cessation of intensive radioactive discharges and the construction of a system of protective water bodies, the doses to aquatic biota noticeably decreased. Current levels of exposure to fish in

  5. Inorganic, radioisotopic and organic analysis of 241-AP-101 tank waste

    Battelle received five samples from Hanford waste tank 241-AP-101, taken at five different depths within the tank. No visible solids or organic layer were observed in the individual samples. Individual sample densities were measured, then the five samples were mixed together to provide a single composite. The composite was homogenized and representative sub-samples taken for inorganic, radioisotopic, and organic analysis. All analyses were performed on triplicate sub-samples of the composite material. The sample composite did not contain visible solids or an organic layer. A subsample held at 10 C for seven days formed no visible solids. The characterization of the 241-AP-101 composite samples included: (1) Inductively-coupled plasma spectrometry for Ag, Al, Ba, Bi, Ca, Cd, Cr, Cu, Fe, K, La, Mg, Mn, Na, Nd, Ni, P, Pb, Pd, Ru, Rh, Si, Sr, Ti, U, Zn, and Zr (Note: Although not specified in the test plan, As, B, Be, Co, Li, Mo, Sb, Se, Sn, Tl, V, W, and Y were also measured and reported for information only) (2) Radioisotopic analyses for total alpha and total beta activities, 3H, 14C, 60Co, 79Se, 90Sr, 99Tc as pertechnetate, 106Ru/Rh, 125Sb, 134Cs, 137Cs, 152Eu, 154Eu, 155Eu, 238Pu, 239+240Pu, 241Am, 242Cm, and 243+244Cm; (3) Inductively-coupled plasma mass spectrometry for 237Np, 239Pu, 240Pu, 99Tc, 126Sn, 129I, 231Pa, 233U, 234U, 235U, 236U, 238U, 241AMU, 242AMU, 243AMU, As, B, Be, Ce, Co, Cs, Eu, I, Li, Mo, Pr, Rb, Sb, Se, Ta, Te, Th, Tl, V, and W; (4) total U by kinetic phosphorescence analysis; (5) Ion chromatography for Cl, F, NO2, NO3, PO4, SO4, acetate, formate, oxalate, and citrate; (6) Density, inorganic carbon and organic carbon by two different methods, mercury, free hydroxide, ammonia, and cyanide. The 241-AP-101 composite met all contract limits (molar ratio of analyte to sodium or ratio of becquerels of analyte to moles of sodium) defined in Specification 7 for Envelope A. Except for a few cases, the characterization results met or surpassed the

  6. Extensions to the Beta Secondary Standard BSS 2

    Behrens, R.; Buchholz, G.

    2011-11-01

    Since several years, the irradiation facility for beta radiation, the Beta Secondary Standard BSS 2 developed at PTB, has been in worldwide use for the performance of irradiations with calibrated beta sources. Due to recent developments in eye tumor therapy, in eye lens dosimetry, and in soft- and hardware technology, several extensions have been added to the BSS 2. These extensions are described in this paper: 1. The possibility of using a 106Ru/106Rh beta source was added as this radionuclide is often used in tumor therapy. 2. The (small) contribution due to photon radiation was included in the dose (rate) reported by the BSS 2, as this was missing in the past. 3. The quantity personal dose equivalent at a depth of 3 mm, Hp(3), was implemented due to recent findings on the radio sensitivity of the eye lens regarding cataract induction and the subsequent lowering of the dose limit from 150 mSv down to 20 mSv per year; 4. The correction for ambient conditions (air temperature, pressure, and relative humidity) was improved in order to adequately handle the quantity Hp(3) and in order to extend the range of use beyond 25°C. 5. A checksum test was added to the software to secure the calibration data against (un)intended changes. 6. The connection of the PC and the BSS 2 has been changed to a network interface (TCP/IP) in order to be able to use up-to-date computers not containing a parallel and a serial port. 7. A rod phantom was added in order to make sure the mechanical set-up is of high quality. All these extensions have been implemented in the PTB's BSS 2 model. The routine implementation of extension 1 is still under investigation by the manufacturer. The commercially available BSS 2 will contain extensions 2 to 6 starting approximately in 2012, while extension 7 has already been incorporated since 2011. Extensions 2 to 4 will also be available for old BSS 2 versions via a software update, starting approximately at the beginning of 2012. Extension 6 will be

  7. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO4, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO4 marked with 106Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO4 is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO4 → RuO2 + O2. To do this, we have tried to introduce gaseous RuO4 into a container placed in an electric oven, and to follow the reaction by γ counting. (author)

  8. Inorganic, radioisotopic and organic analysis of 241-AP-101 tank waste

    SK Fiskum; PR Bredt; JA Campbell; LR Greenwood; OT Farmer; GJ Lumetta; GM Mong; RT Ratner; CZ Soderquist; RG Swoboda; MW Urie; JJ Wagner

    2000-06-28

    Battelle received five samples from Hanford waste tank 241-AP-101, taken at five different depths within the tank. No visible solids or organic layer were observed in the individual samples. Individual sample densities were measured, then the five samples were mixed together to provide a single composite. The composite was homogenized and representative sub-samples taken for inorganic, radioisotopic, and organic analysis. All analyses were performed on triplicate sub-samples of the composite material. The sample composite did not contain visible solids or an organic layer. A subsample held at 10 C for seven days formed no visible solids. The characterization of the 241-AP-101 composite samples included: (1) Inductively-coupled plasma spectrometry for Ag, Al, Ba, Bi, Ca, Cd, Cr, Cu, Fe, K, La, Mg, Mn, Na, Nd, Ni, P, Pb, Pd, Ru, Rh, Si, Sr, Ti, U, Zn, and Zr (Note: Although not specified in the test plan, As, B, Be, Co, Li, Mo, Sb, Se, Sn, Tl, V, W, and Y were also measured and reported for information only) (2) Radioisotopic analyses for total alpha and total beta activities, {sup 3}H, {sup 14}C, {sup 60}Co, {sup 79}Se, {sup 90}Sr, {sup 99}Tc as pertechnetate, {sup 106}Ru/Rh, {sup 125}Sb, {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 154}Eu, {sup 155}Eu, {sup 238}Pu, {sup 239+240}Pu, {sup 241}Am, {sup 242}Cm, and {sup 243+244}Cm; (3) Inductively-coupled plasma mass spectrometry for {sup 237}Np, {sup 239}Pu, {sup 240}Pu, {sup 99}Tc, {sup 126}Sn, {sup 129}I, {sup 231}Pa, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 241}AMU, {sup 242}AMU, {sup 243}AMU, As, B, Be, Ce, Co, Cs, Eu, I, Li, Mo, Pr, Rb, Sb, Se, Ta, Te, Th, Tl, V, and W; (4) total U by kinetic phosphorescence analysis; (5) Ion chromatography for Cl, F, NO{sub 2}, NO{sub 3}, PO{sub 4}, SO{sub 4}, acetate, formate, oxalate, and citrate; (6) Density, inorganic carbon and organic carbon by two different methods, mercury, free hydroxide, ammonia, and cyanide. The 241-AP-101 composite met all

  9. Reference beta radiations for calibrating dosemeters and dose ratemeters and for determining their response as a function of beta radiation energy. 1. ed.

    This International Standard specifies the requirements for reference beta radiations produced by radionuclide sources to be used for the calibration of protection level dosemeters and dose ratemeters, and for the determination of their response as a function of beta energy. It gives the characteristics of radionuclides which have been used to produce reference beta radiations, gives examples of suitable source constructions and describes methods for the measurement of the residual maximum beta energy and the absorbed dose rate at a depth of 7 mg·cm-2 in a semi-infinite tissue-equivalent medium. The energy range involved lies between 66 keV and 3.6 MeV and the absorbed dose rates are in the range from about 10 μGy·h-1 (1 mrad·h-1) to at least 10 Gy·h-1 (103 rad·h-1). This International Standard proposes two series of beta reference radiations from which the radiation necessary for determining the characteristics (calibration and energy response) of an instrument shall be selected. Series 1 reference radiations are produced by radionuclide sources used with beam flattening filters designed to give uniform dose rates over a large area at a specific distance. The proposed sources of 90Sr+90Y, 204TI and 147Pm produce maximum dose rates of approximately 5mGy·h-1 (0.5 rad·h-1). Series 2 reference radiations are produced without the use of beam flattening filters which allows a range of source-to-calibration plane distances to be used. Close to the sources only relatively small areas of uniform dose rate are produced but this Series has the advantage of extending the energy and dose rate ranges beyond those of Series 1. The radionuclides used are those of Series 1 with the addition of the radionuclides 14C and 106Ru+106Rh; these sources produce dose rates of up to 10 Gy·h-1 (103 rad·h-1)

  10. Results of 1999 Spectral Gamma-Ray and Neutron Moisture Monitoring of Boreholes at Specific Retention Facilities in the 200 East Area, Hanford Site

    Twenty-eight wells and boreholes in the 200 East Are% Hanford Site, Washington were monitored in 1999. The monitored facilities were past-practice liquid waste disposal facilities and consisted of six cribs and nineteen ''specific retention'' cribs and trenches. Monitoring consisted of spectral gamma-ray and neutron moisture logging. All data are included in Appendix B. The isotopes 137Cs, 60Co, 235U, 238U, and 154Eu were identified on spectral gamma logs from boreholes monitoring the PUREX specific retention facilities; the isotopes 137Cs, 60Co, 125Sb, and 154Eu were identified on the logs from boreholes at the BC Controlled Area cribs and trenches; and 137Cs, 60Co, and 125Sb were, identified on the logs from boreholes at the BX specific retention trenches. Three boreholes in the BC Controlled Area and one at the BX trenches had previous spectral gamma logs available for comparison with 1999 logs. Two of those logs showed that changes in the subsurface distribution of 137CS and/or 60Co had occurred since 1992. Although the changes are not great, they do point to continued movement of contaminants in the vadose zone. The logs obtained in 1999 create a larger baseline for comparison with future logs. Numerous historical gross gamma logs exist from most of the boreholes logged. Qualitative comparison of those logs with the 1999 logs show many substantial changes, most of which reflect the decay of deeper short-lived isotopes, such as 106Ru and 125Sb, and the much slower decay of shallower and longer-lived isotopes such as 137Cs. The radionuclides 137Cs and 60Co have moved in two boreholes since 1992. Given the amount of movement and the half-lives of the isotopes, it is expected that they will decay to insignificant amounts before reaching groundwater. However, gamma ray logging cannot detect many of the contaminants of interest such as 99Tc, NO3, or 129I, all of which can be highly mobile in the vadose zone and, for the radionuclides, have long half-lives

  11. Department of Radiation Shielding and Dosimetry: Overview

    Full text: The research activities of the Department in 1999, similarly to the previous year were focused on the following problems: -Dosimetry for medical purposes, - Microdosimetry at the nanometre level, -Numerical modelling of interaction of radiation with matter. The following activities should be emphasized: - DOSIMETRY: The method for standardisation of a scintillation detector with NE102A organic scintillator in the terms of absorbed dose has been accomplished. The method is based on the use of small size ionisation chamber ''pipe type'' with sensitive area of 0.8 cm2. The response of the chamber has been traced in SSDL laboratory against secondary standards. This scintillation detector has been used for standardisation of the of the absorbed dose depth dependence in water for the 106Ru ophthalmic applicators. A new type of an ionisation chamber, called Ring Ionisation Chamber for standardising of absorbed dose from beta-radioactive wires, used for endovascular brachytherapy has been designed. This activity is supported by Grant KBN Nr 4P05C01417. -MICRODOSIMETRY: After a prolonged time of research the success has ben achieved in developing the method for measuring the ion cluster spectra at the nanometre level. The ion clusters spectra created along nanometre size track of alpha particles with energy of 4.6 MeV were measured with the Jet Counter set up. The ion cluster spectra of nanometre size with dimension ranging from 0.15 to 13 nm (at unit density scale) have been measured. The deconvolution method, converting the measured spectra to the true ones has been developed. The results are the first of this kind ever obtained. This activity was supported by IV CEC Framework as well as by Polish Commission for Scientific Research. - NUMERICAL MODELLING: MCNP-A General Monte Carlo N-Particle Transport Code was used for modelling electron penetration in water eye ball phantom. Depth dose distributions in eye phantom have been calculated for different shapes of

  12. Reconstruction of radionuclide contamination of the Techa River caused by liquid waste discharge from radiochemical production at the Mayak Production Association

    , rather than relying on a more uncertain reconstruction of quantities released at the point of discharge. Radionuclides considered include 90Sr, 106Ru, 137Cs, and 144Ce. Estimated concentrations of selected radionuclides at various times are presented

  13. Reconstruction of radionuclide contamination of the Techa River caused by liquid waste discharge from radiochemical production at the Mayak Production Association

    Mokrov, Y.; Glagolenko, Y.; Napier, B.

    2000-07-01

    river, rather than relying on a more uncertain reconstruction of quantities released at the point of discharge. Radionuclides considered include {sup 90}Sr, {sup 106}Ru, {sup 137}Cs, and {sup 144}Ce. Estimated concentrations of selected radionuclides at various times are presented.

  14. Reconstruction of radionuclide contamination of the Techa River caused by liquid waste discharge from radiochemical production at the Mayak Production Association.

    Mokrov, Y; Glagolenko, Y; Napier, B

    2000-07-01

    than relying on a more uncertain reconstruction of quantities released at the point of discharge. Radionuclides considered include 90Sr, 106Ru, 137Cs, and 144Ce. Estimated concentrations of selected radionuclides at various times are presented. PMID:10855774

  15. Non destructive determination of fuel burn up of the RA reactor at Vinca by gamma radiation spectra analysis

    The problem of non destructive determination of the burn-up of used up fuel of the powerful experimental reactor RA at Vinca, by the analysis of gamma radiation spectra using a gamma-spectrometer with a semiconductor Ge(Li) detector has been studied. The first part of this analytical problem is concerned with calculation of fuel burn up. In the preparation of its solution material and energetic balance of the fuel composition and conditions of fuel irradiation. The obtained solution uses numerical methods. In this solution fuel burn up is determined: 1) on the basis of the data on the composition of 106Ru, 134Cs and 137Cs, than 2) from a series of the data on the fuel and reactor and 3) on the basis of those, numerous, literature data which participate in defining of the balance of fuel burn up process. The second part of this problem refers to determination of the composition of the above gamma radioactive fission products from the obtained instrumental spectrum. Under strictly defined conditions of the measurement of gamma radiation from the fuel elements, when a determined type of the gamma ray collimator is used, the photo peak area of the corresponding line in the instrumental spectrum is defined as a function of the fission products activity, energy and yield of its gamma rays, the thickness of fuel and added absorbers as well as the dimensions of the collimator used. On this basis, the activity quotient of two fission products is a function of: 1) two areas of two photo peaks of their two lines and 2) some of the above cited values. Unknown magnitudes of the remaining values of the same relation are determined from the photo peak areas of the lines of the complex spectrum of one fission product by solving a system of equations. Accuracy of the solutions of these two separate parts of the observed analytical problems was confirmed experimentally. The results obtained are characterized by high repeatability. Total errors are generally greater, primarily due to

  16. Results For The Third Quarter 2009 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminant Results

    This report details the chemical and radionuclide contaminant results for the characterization of the 2009 Third Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC). Information from this characterization will be used by Liquid Waste Operations (LWO) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System. Recently, a review of the radionuclide inventory in Saltstone Vaults 1 and 4 identified several additional radionuclides, not currently in the WAC, which require quantification (40K, 108mAg, 133Ba, 207Bi, 227Ac, 228Ra, 228Th, 231Pa, 247Cm, 249Cf, 251Cf). In addition, several of the radionuclides previously reported with minimum detection limits below the requirements listed in the WAC required analysis with reduced detection limits to support future inventory reporting requirements (22Na, 26Al, 59Ni, 94Nb, 106Ru, 144Ce, 152Eu, 155Eu, 226Ra). This added scope was formally requested in a revision to the standing Technical Task Request for CY2009 Saltstone support and is further discussed in several supporting documents. The following conclusions are drawn from the analytical results provided in this report: (1) The concentrations of the reported chemical and radioactive contaminants are less than their respective WAC targets or limits unless noted in this section. (2) The reported detection limits for 59Ni, 94Nb, 247Cm, and 249Cf are above the limits requested by LWO; however, they are below the achievable limits established by Analytical Development (AD). (3) The reported detection limit of isopropanol is lower than its WAC Limit for accident analysis in Appendix 8.1, but higher than its WAC concentration given in Table 4 for vault flammability. The higher detection limit is expected based on current analytical capabilities and is documented in the Task

  17. Nuclear fuel cycle and marine environment. Behavior of the Rhone river effluents in the mediterranean sea and of wastes dumped in the northeast atlantic; Cycle du combustible nucleaire et milieu marin. Devenir des effluents rhodaniens en mediterranee et des dechets immerges en atlantique nord-est

    Charmasson, S

    1998-07-01

    Man-made radionuclides released into the marine environment by the installations from the nuclear fuel cycle are used as tracers of various bio-geochemical processes. Several installations belonging to the whole nuclear fuel cycle, except the uranium mining, are set up on the Rhone River Banks. The sea disposal of low and intermediate level radioactive waste has never been authorized in the Mediterranean sea but several sites have been used in the North-East especially in abyssal waters. Radionuclides released by the Rhone river installations are used in order to study the dynamics of the Rhone inputs into the Mediterranean Sea. In the river, freshwater samples reflect quite accurately the discharge composition with a predominance of {sup 106}Ru, a radionuclide mostly released by the spent nuclear fuel reprocessing plant in Marcoule. Conversely, at the Rhone mouth, in the sediment compartment {sup 106}Ru yields to caesium isotopes ({sup 134}Cs and {sup 137}Cs) in importance. As these two isotopes demonstrate very different half-lives (30,2 and 2,1 years respectively), the temporal evolution of their ratio acts as a chronometer enabled to date sediment accumulation near the river mouth. Mean accumulation rates greater than 35 cm y{sup -1} have been determined in the pro-deltaic zone near the Roustan buoys over the period 1983-1991. Accumulation rates decrease rapidly with distance from the mouth and therefore most of the {sup 137}Cs inventory in this part of the Gulf of Lions is limited to the pro-deltaic area. A first study about the part the different {sup 137}Cs sources in the Mediterranean Sea play in this inventory has been carried out. Direct (atmospheric) and indirect (fluviatile) inputs due to fallout from both past nuclear tests and the Chernobyl accident could contribute to this inventory at the highest to 40 % while the industrial releases could contribute at the lowest to 60 %. The last site used for the dumping of low and intermediate level radioactive

  18. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide; Volatilite du ruthenium au cours des operations de vitrification des produits de fission. 1. partie distillation des solutions nitriques calcination des concentrats 2. partie fixation sur un tube d'acier decomposition du peroxyde

    Ortins de Bettencourt, A.; Jouan, A. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO{sub 4}, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO{sub 4} marked with {sup 106}Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO{sub 4} is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO{sub 4} {yields} RuO{sub 2} + O{sub 2}. To do this, we have tried to introduce gaseous RuO{sub 4} into a container placed in an electric oven, and to follow the reaction by {gamma} counting. (author) [French] Au cours de la vitrification des solutions de produits de

  19. Nuclear fuel cycle and marine environment. Behavior of the Rhone river effluents in the mediterranean sea and of wastes dumped in the northeast atlantic

    Man-made radionuclides released into the marine environment by the installations from the nuclear fuel cycle are used as tracers of various bio-geochemical processes. Several installations belonging to the whole nuclear fuel cycle, except the uranium mining, are set up on the Rhone River Banks. The sea disposal of low and intermediate level radioactive waste has never been authorized in the Mediterranean sea but several sites have been used in the North-East especially in abyssal waters. Radionuclides released by the Rhone river installations are used in order to study the dynamics of the Rhone inputs into the Mediterranean Sea. In the river, freshwater samples reflect quite accurately the discharge composition with a predominance of 106Ru, a radionuclide mostly released by the spent nuclear fuel reprocessing plant in Marcoule. Conversely, at the Rhone mouth, in the sediment compartment 106Ru yields to caesium isotopes (134Cs and 137Cs) in importance. As these two isotopes demonstrate very different half-lives (30,2 and 2,1 years respectively), the temporal evolution of their ratio acts as a chronometer enabled to date sediment accumulation near the river mouth. Mean accumulation rates greater than 35 cm y-1 have been determined in the pro-deltaic zone near the Roustan buoys over the period 1983-1991. Accumulation rates decrease rapidly with distance from the mouth and therefore most of the 137Cs inventory in this part of the Gulf of Lions is limited to the pro-deltaic area. A first study about the part the different 137Cs sources in the Mediterranean Sea play in this inventory has been carried out. Direct (atmospheric) and indirect (fluviatile) inputs due to fallout from both past nuclear tests and the Chernobyl accident could contribute to this inventory at the highest to 40 % while the industrial releases could contribute at the lowest to 60 %. The last site used for the dumping of low and intermediate level radioactive waste in the North-East Atlantic has been

  20. Nuclear fuel cycle and marine environment. Behavior of the Rhone river effluents in the mediterranean sea and of wastes dumped in the northeast atlantic; Cycle du combustible nucleaire et milieu marin. Devenir des effluents rhodaniens en mediterranee et des dechets immerges en atlantique nord-est

    Charmasson, S

    1998-07-01

    Man-made radionuclides released into the marine environment by the installations from the nuclear fuel cycle are used as tracers of various bio-geochemical processes. Several installations belonging to the whole nuclear fuel cycle, except the uranium mining, are set up on the Rhone River Banks. The sea disposal of low and intermediate level radioactive waste has never been authorized in the Mediterranean sea but several sites have been used in the North-East especially in abyssal waters. Radionuclides released by the Rhone river installations are used in order to study the dynamics of the Rhone inputs into the Mediterranean Sea. In the river, freshwater samples reflect quite accurately the discharge composition with a predominance of {sup 106}Ru, a radionuclide mostly released by the spent nuclear fuel reprocessing plant in Marcoule. Conversely, at the Rhone mouth, in the sediment compartment {sup 106}Ru yields to caesium isotopes ({sup 134}Cs and {sup 137}Cs) in importance. As these two isotopes demonstrate very different half-lives (30,2 and 2,1 years respectively), the temporal evolution of their ratio acts as a chronometer enabled to date sediment accumulation near the river mouth. Mean accumulation rates greater than 35 cm y{sup -1} have been determined in the pro-deltaic zone near the Roustan buoys over the period 1983-1991. Accumulation rates decrease rapidly with distance from the mouth and therefore most of the {sup 137}Cs inventory in this part of the Gulf of Lions is limited to the pro-deltaic area. A first study about the part the different {sup 137}Cs sources in the Mediterranean Sea play in this inventory has been carried out. Direct (atmospheric) and indirect (fluviatile) inputs due to fallout from both past nuclear tests and the Chernobyl accident could contribute to this inventory at the highest to 40 % while the industrial releases could contribute at the lowest to 60 %. The last site used for the dumping of low and intermediate level radioactive

  1. Study of solution speciation, soil retention and soil-plant transfer of zirconium; Etude de la speciation en solution, de la retention dans les sols et du transfert sol-plante du zirconium

    Ferrand, E

    2005-12-15

    Within the framework of the risks prevention policy of Andra, the radioactive zirconium introduction ({sup 93}Zr and {sup 95}Zr) into the environment could be carried out starting from the nuclear waste whose storage is envisaged in deep geological layers. Thus, the goal of this study was to evaluate the parameters and phenomena influencing speciation (various chemical forms) and the soil-plant transfer of zirconium. Experiments of adsorption/desorption of zirconium with different ligands likely to be present in soils (goethite and humic acid) and with two soils, with contrasted characteristics, close to the underground research laboratory of Andra (Meuse) were carried out. These results of adsorption were then confronted with those obtained by the MUSIC and NICA-DONNAN models carried out using the computer code ECOSAT. Zr presents a strong affinity for the two types of soils and the soils constituents. Specific interactions of internal sphere type with the goethite were highlighted using the model. Soil-solution partition coefficients, or K{sub d}, values increase with pH and contact time. Various types of edible plants, pea (Pisum sativum L.) and tomato (Lycopersicon esculentum L cv. St Pierre) were cultivated in hydroponic conditions and in soils spiked with various sources of Zirconium. The maximum zirconium contents are mainly measured in the roots of the plants. The soil-plant transfer factors measured during these experiments show a weak bioavailability of zirconium. An influence of speciation on Zr bioavailability is however highlighted. Some chemical forms, such as oxychloride or acetate, are more easily mobilized than others by the plant. (author)

  2. Vertical distribution and estimated doses from artificial radionuclides in soil samples around the Chernobyl nuclear power plant and the Semipalatinsk nuclear testing site.

    Yasuyuki Taira

    Full Text Available For the current on-site evaluation of the environmental contamination and contributory external exposure after the accident at the Chernobyl Nuclear Power Plant (CNPP and the nuclear tests at the Semipalatinsk Nuclear Testing Site (SNTS, the concentrations of artificial radionuclides in soil samples from each area were analyzed by gamma spectrometry. Four artificial radionuclides ((241Am, (134Cs, (137Cs, and (60Co were detected in surface soil around CNPP, whereas seven artificial radionuclides ((241Am, (57Co, (137Cs, (95Zr, (95Nb, (58Co, and (60Co were detected in surface soil around SNTS. Effective doses around CNPP were over the public dose limit of 1 mSv/y (International Commission on Radiological Protection, 1991. These levels in a contaminated area 12 km from Unit 4 were high, whereas levels in a decontaminated area 12 km from Unit 4 and another contaminated area 15 km from Unit 4 were comparatively low. On the other hand, the effective doses around SNTS were below the public dose limit. These findings suggest that the environmental contamination and effective doses on the ground definitely decrease with decontamination such as removing surface soil, although the effective doses of the sampling points around CNPP in the present study were all over the public dose limit. Thus, the remediation of soil as a countermeasure could be an extremely effective method not only for areas around CNPP and SNTS but also for areas around the Fukushima Dai-ichi Nuclear Power Plant (FNPP, and external exposure levels will be certainly reduced. Long-term follow-up of environmental monitoring around CNPP, SNTS, and FNPP, as well as evaluation of the health effects in the population residing around these areas, could contribute to radiation safety and reduce unnecessary exposure to the public.

  3. Comparison of two selective separation method for {sup 93}Zr by using TRU and TEVA resins

    Oliveira, Thiago C.; Oliveira, Arno Heeren de, E-mail: tco@cdtn.b, E-mail: heeren@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear; Monteiro, Roberto Pellacani G., E-mail: rpgm@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The zirconium isotope {sup 93}Zr is a long-lived pure {beta}-particle-emitting radionuclide produced from {sup 235}U fission and from neutron activation of the stable isotope {sup 92}Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, {sup 93}Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of {sup 93}Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. The aim of this work was to compare two radiochemical procedure based on selective extraction using an anion-exchange chromatography, TRU and TEVA resins, in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique. To set up the radiochemical separation procedure for zirconium, a tracer solution of {sup 95}Zr and its 724.19 keV {gamma}-ray measurements by {gamma} - spectrometry were used in order to follow the behavior of zirconium during the radiochemical separation. A tracer solution of {sup 55}Fe, the main interference in the LSC measurements, was used in order to verify the decontamination factor during the separation process. The limit of detection of the 0.05 Bq 1{sup -1} was obtained for {sup 55}Fe standard solutions by using a sample:cocktail ratio of 3:17 mL for Optiphase Hisafe 3 cocktail. (author)

  4. Actual problems of the distribution of radioisotopes and the kinetics of their excretion

    The paper deals with the character of the distribution of radioisotopes (Cs, Rb, Sr, Ca, Ba, Ra, Ce, La, Pm, Sb, Te, Pu, Y, Zr, Ru, Nb, Po), depending on the way and rhythm of their administration, the physiological state of the organism and physico-chemical factors. It was shown that 10 to 18% of some radioisotopes (Cs137, Nb95, Zr95, Y91, Ru106, and Ce144) were retained in the lungs when administered by inhalation. Caesium and especially strontium are absorbed from the lungs extremely rapidly, while the absorption of ruthenium, cerium and particularly plutonium is slow. The type of distribution of radioactive isotopes does not depend on the species of the animal, as is sometimes the case with the kinetics of their elimination. The rate of Ce144 removal from canine and feline liver is exceedingly low, unlike that in rats, m ice, guinea pigs and rabbits. The amount of the deposited material and the kinetics of its elimination in case of Sr90 and Pu239 are determined by the rhythm and duration of the isotope intake. From the data presented it can be seen that the results of experiments with a single injection do not always characterize the kinetics of the accumulation of the isotope in cases of its continuous administration. Chelate compounds (EDTA, hexametaphosphate - HMP), isotope and non-isotope carriers, and the pH of the original solution change the distribution of hydrolysed elements (cerium, yttrium) and do not affect the type of distribution of elements that are highly-soluble in water. When the pH of the original solution increases, the amount of cerium and yttrium deposition also increases in the organs containing much reticuloendothelial elements (liver, spleen) and decreases in the skeleton and kidneys. The acceleration of the Ce144 and Y91 excretion caused by the chelate compounds (EDTA, HMP) is delayed with the increase of the pH of the original solution. (author)

  5. Lichen (sp. Cladonia) as a deposition indicator for transuranium elements investigated with the Chernobyl fallout

    Paatero, Jussi; Jaakkola, Timo; Kulmala, Seija [Helsinki Univ., Radiochemistry Lab., Helsinki (Finland)

    1998-06-01

    The feasibility of employing carpet-forming lichens (sp. Cladonia) as a measure for the deposition of transuranium elements was investigated with the Chernobyl fallout. In Finland, the deposition of these elements after the accident was very uneven. The highest deposition values for {sup 238}Pu, {sup 239,240}Pu, {sup 241}Am, {sup 242}Cm and {sup 243,244}Cm were 5.7, 3.0, 1.3, 98 and 0.025 Bq m{sup -2}, respectively. The amount of deposited {sup 239,240}Pu was, however, only some percent of the fallout of the nuclear test explosions of the 1950s and the 1960s. Instead, practically no {sup 242}Cm was released into the environment during the weapons` testing. The correlation between the refractory nuclides {sup 238}Pu, {sup 239,240}Pu, {sup 241}Am, {sup 242}Cm, {sup 95}Zr and {sup 144}Ce in lichen was high (r = 0.709-0.979), but the correlation between the transuranium elements and volatile {sup 137}Cs was much lower (r = 0.227-0.276). The calculated biological half-lives of Pu and Am in lichen were 730 and 320 days, respectively. The {sup 238}Pu/{sup 239,240}Pu activity ratio in the top parts of lichen samples was 0.54 ``+`` 0.02 corresponding to a burn-up value of 12 MWd kg{sup -1}. The {sup 238}Pu/{sup 239,240}Pu activity ratios of 0.43 and 0.69 measured from two isolated hot particles correspond to burn-up values of 10 and 14 MWd kg{sup -1}, respectively. (author).

  6. Radioactive effluents in Savannah River. Summary report for 1992

    Winn, W.G.

    1993-09-21

    During 1992, the radioactive effluents in the Savannah River were less than those observed in 1991. Vogtle reported no significant releases in 1992, and in earlier years Vogtle improvements in pre-processing their releases had already effected a decreasing trend in release levels. Their effluents continue to be dominated by {sup 58}Co, which had a maximum concentration of only 0.068 pCi/L, which is just 1/3 of the maximum observed in 1991. Many of the other man-made radionuclides observed in earlier years have now decreased to where some are not even detected, and no new radionuclides were detected in the 1992 Vogtle effluents. In addition to {sup 58}Co, low levels of {sup 60}Co were frequently observed, but only traces of {sup 54}Mn and {sup 95}Nb were observed. Contrary to earlier years no {sup 51}Cr, {sup 57}Co, {sup 59}Fe, or {sup 95}Zr were seen in 1992. Tritium and {sup 137}Cs were also monitored, but their levels generally remain consistent with known SRS sources. The maximum tritium observed near Vogtle was 2 pCi,/mL. The maximum downstream tritium was higher (3.8 pCi/mL), primarily due to the tritium release from K-Reactor in December 1991; however, the levels had abated significantly prior to collection of the tritium samples of the present study. In addition to natural sources, the general levels in the Savannah River are due to routine releases from the effluent treatment facility and seepage basin migration into plant streams that flow into the river.

  7. Commercial testing of a unit for high-temperature removal of corrosion products from water coolant

    Recently interest has been shown in the Soviet and foreign nuclear power industries in the removal of corrosion products from hot (150-300C) flows of water coolant. This interest is explained by the need to increase the efficiency of removal of corrosion products from the loops of an atomic power plant in order to lower the rate of formation of deposits of such corrosion products on the in-loop surfaces, especially the heat-transfer surfaces. Filter materials - heat-resistant titanium-based inorganic sorbents - have been developed and investigated for removal of corrosion products from hot flows of water coolant at atomic power plants. The basic regularities underlying the process of filtration on sorbent beds have been studied. The efficiency of removal from water of corrosion products and radionuclides produced by corrosion is 50-95% for total iron, 80-85% for 51Cr, 90% for 54Mn, 99% for 56Mn, 90% for 59Fe, 50-70% for 60Co, 95% for 95Nb, and 95% for 95Zr. Technologies for sorbent regeneration and hydraulic reloading, a design for a high-temperature filter, and the layout of a treatment system have been developed. The information amassed to date served as the basis for the creation of the high-temperature treatment unit at the Beloyarsk Atomic Power Plant. This paper presents the results of the testing of a commercial unit for high-temperature treatment of water coolant. Such units have been included in the plans for a number of atomic power plants and heat-and-electric generating plants. The testing therefore was conducted to gain experience in their operation

  8. Radioactive effluents in Savannah River. Summary report for 1992

    During 1992, the radioactive effluents in the Savannah River were less than those observed in 1991. Vogtle reported no significant releases in 1992, and in earlier years Vogtle improvements in pre-processing their releases had already effected a decreasing trend in release levels. Their effluents continue to be dominated by 58Co, which had a maximum concentration of only 0.068 pCi/L, which is just 1/3 of the maximum observed in 1991. Many of the other man-made radionuclides observed in earlier years have now decreased to where some are not even detected, and no new radionuclides were detected in the 1992 Vogtle effluents. In addition to 58Co, low levels of 60Co were frequently observed, but only traces of 54Mn and 95Nb were observed. Contrary to earlier years no 51Cr, 57Co, 59Fe, or 95Zr were seen in 1992. Tritium and 137Cs were also monitored, but their levels generally remain consistent with known SRS sources. The maximum tritium observed near Vogtle was 2 pCi,/mL. The maximum downstream tritium was higher (3.8 pCi/mL), primarily due to the tritium release from K-Reactor in December 1991; however, the levels had abated significantly prior to collection of the tritium samples of the present study. In addition to natural sources, the general levels in the Savannah River are due to routine releases from the effluent treatment facility and seepage basin migration into plant streams that flow into the river

  9. Radiation and migration characteristics of the Chernobyl radionuclides causing environmental contamination in various provinces in the European sector of the USSR

    The radionuclide composition of the Chernobyl fallout has been studied over practically the whole European sector of the Union of Soviet Socialist Republics using soil sampling methods, the upper turf layer of the soil being treated as a natural planchet. Regular observation of the migration of radionuclides in the soil-plant system has been in progress since July 1986 at a stationary network of landscape-geochemical testing sites and sampling grounds; two types of natural ecosystem have been looked at - meadowland and forest. The behaviour of individual radionuclides is evaluated using the coefficient of fractionation in relation to 95Zr or 144Ce. The migration characteristics of individual radionuclides were revealed to depend both on soil conditions and on the type of radionuclide composition. The main features of the vertical profile of radionuclide concentrations in soils were clear 2-3 months after the contamination occurred. During subsequent years the migrational profiles changed depending on the individual characteristics of the radionuclides and soil conditions. Information on the radionuclide composition and evolution of the vertical distribution of radionuclides in soils was used to calculate and predict exposure gamma radiation dose from the Chernobyl fallout in various provinces in the European sector of the Soviet Union. When monitoring forest sampling grounds, samples were taken of the soil (litter, humus, mineral layer), the phytomasses of ligneous species (pine needles, leaves, bark, wood), moss and lichens. The main aim of monitoring surveys from 1986 to 1989 was to evaluate the distribution of Chernobyl radionuclides in the compartments of the forest ecosystem. In particular, 60-80% of the total amount of 137Cs was bound with the humus layer of the forest soil. 4 refs, 6 figs, 3 tabs

  10. Modeling activity transport in the CANDU heat transport system

    The release and transport of corrosion products from the surfaces of primary coolant system components is a serious concern for all water-cooled nuclear power plants. The consequences of high levels of corrosion product transport are twofold: a) increased corrosion product (crud) deposition on fuel cladding surfaces, leading to reduced heat transfer and the possibility of fuel failures, and b) increased production of radioactive species by neutron activation, resulting in increased out-of-core radiation fields and worker dose. In recent years, a semi-empirical activity transport model has been successfully developed to predict the deposition of radionuclides, including 60Co, 95Zr, 124Sb and fission products, around the CANDU® primary Heat Transport System (HTS), and to predict radiation fields at the steam generators and reactor face. The model links corrosion of the carbon steel outlet feeders to magnetite and radionuclide deposition on steam generator and inlet piping surfaces. This paper will describe the model development, key assumptions, required inputs, and model validation. The importance of reactor artefact characterization in the model development will be highlighted, and some key results will be presented, including oxide morphology and loadings, and radionuclide distributions within the oxide. The predictive capabilities of the model will also be described, including predictions of oxide thickness and the effects of changes in chemistry parameters such as alkalinity. While the model was developed primarily for the CANDU® HTS, the information gained during model development regarding corrosion product and radionuclide transport and deposition can also provide insights into activity transport in other water-cooled reactor systems. (author)

  11. Measurement and analysis of JOYO MK-II spent MOX fuel decay heat (3)

    It is important to precisely evaluate decay heat of the spent MOX fuel not only from the viewpoint of reactor safety concerning a decay heat removal at the reactor shut down, but also for thermal design of the spent fuel storage and handling facility. In order to obtain the experimental data and to improve the accuracy of calculation, the decay heat of spent fuel subassemblies (Burn-up: 66 GWd/t) of the JOYO MK-II core was measured from the cooling time of 319 to 729 days. Measured decay heat of the spent fuel subassemblies was 351±16∼158±9W. The low flow rate measurement to make large temperature gradient between inlet and outlet of measurement system has allowed us to measure decay heat after cooling time of 500 days by less than 6% error (1σ). The decay heat was calculated by 'ORIGEN2' code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. Then it was compared with the measured value. The ratios between calculated and experimental value, C/Es, were approximately between 0.96 and 0.90. There exists a systematic discrepancy (6∼8%) between calculation and measurement, which was larger than the experimental error (1σ = 1.7∼6.0%). Decay heat was generated from actinides and fission products (FP). Major heat sources in actinides are 242Cm, 238Pu and 241Am. However, nuclides except for 242Cm were not considered as the cause of systematic discrepancy because these decay heat was almost constant throughout this measurement. After cooling time of 100 days, 95Zr, 95Nb, 106Rh and 144Pr remained as major FP decay heat source. As as result, the discrepancy appears due to the decay heat calculation uncertainty of 242Cm and these FP, or measurement error and further investigation will be required. (author)

  12. Study of solution speciation, soil retention and soil-plant transfer of zirconium

    Within the framework of the risks prevention policy of Andra, the radioactive zirconium introduction (93Zr and 95Zr) into the environment could be carried out starting from the nuclear waste whose storage is envisaged in deep geological layers. Thus, the goal of this study was to evaluate the parameters and phenomena influencing speciation (various chemical forms) and the soil-plant transfer of zirconium. Experiments of adsorption/desorption of zirconium with different ligands likely to be present in soils (goethite and humic acid) and with two soils, with contrasted characteristics, close to the underground research laboratory of Andra (Meuse) were carried out. These results of adsorption were then confronted with those obtained by the MUSIC and NICA-DONNAN models carried out using the computer code ECOSAT. Zr presents a strong affinity for the two types of soils and the soils constituents. Specific interactions of internal sphere type with the goethite were highlighted using the model. Soil-solution partition coefficients, or Kd, values increase with pH and contact time. Various types of edible plants, pea (Pisum sativum L.) and tomato (Lycopersicon esculentum L cv. St Pierre) were cultivated in hydroponic conditions and in soils spiked with various sources of Zirconium. The maximum zirconium contents are mainly measured in the roots of the plants. The soil-plant transfer factors measured during these experiments show a weak bioavailability of zirconium. An influence of speciation on Zr bioavailability is however highlighted. Some chemical forms, such as oxychloride or acetate, are more easily mobilized than others by the plant. (author)

  13. Zero Time of Transitory Nuclear Events Derived by Parent-Daughter Systems

    The detection and identification of a nuclear event that results in the dissemination of radioactive products into the environment can be realized by dating the age of the event. In order to correct observed activities for the decay since the occurrence of the event, the age must be known to a high level of confidence. Previous papers. described the method to date the age of a nuclear event by measuring the activity of two fission products, which constitute the clock in this application. Within the proficiency test programme for radionuclide laboratories supporting the CTBT, a simulated gamma spectrum with the characteristics of an atmospheric test of a Chinese thermonuclear device, was used to determine the zero time by calculating the theoretical peak area ratio of 95Nb/95Zr. Their approach used only the main gamma lines at 766 and 757 keV and assigned the same detection efficiency to both these close lines. Their methodology of calculating the uncertainty of zero time is subject to comments because it takes the sum of two components (nuclide ratio and activity ratio as function of time) in quadrature. In another paper, the activity of 95Nb as a function of time was presented without any development or expression for the zero time. Analytical equations for zero time and the associated uncertainty calculations were derived in a recent paper using a measured activity ratio of two nuclides and illustrating the procedure by data from the Chinese test. The evaluation of the zero time uncertainty was performed by a very large set of very complicated analytical equations. The present paper aims at developing a procedure to determine the zero time and its uncertainty in a transitory nuclear event by treating a parent-daughter system of 3 nuclides, where one daughter feeds the other one, in addition to its direct feeding by the decay of the parent

  14. Vertical distribution and estimated doses from artificial radionuclides in soil samples around the Chernobyl nuclear power plant and the Semipalatinsk nuclear testing site.

    Taira, Yasuyuki; Hayashida, Naomi; Tsuchiya, Rimi; Yamaguchi, Hitoshi; Takahashi, Jumpei; Kazlovsky, Alexander; Urazalin, Marat; Rakhypbekov, Tolebay; Yamashita, Shunichi; Takamura, Noboru

    2013-01-01

    For the current on-site evaluation of the environmental contamination and contributory external exposure after the accident at the Chernobyl Nuclear Power Plant (CNPP) and the nuclear tests at the Semipalatinsk Nuclear Testing Site (SNTS), the concentrations of artificial radionuclides in soil samples from each area were analyzed by gamma spectrometry. Four artificial radionuclides ((241)Am, (134)Cs, (137)Cs, and (60)Co) were detected in surface soil around CNPP, whereas seven artificial radionuclides ((241)Am, (57)Co, (137)Cs, (95)Zr, (95)Nb, (58)Co, and (60)Co) were detected in surface soil around SNTS. Effective doses around CNPP were over the public dose limit of 1 mSv/y (International Commission on Radiological Protection, 1991). These levels in a contaminated area 12 km from Unit 4 were high, whereas levels in a decontaminated area 12 km from Unit 4 and another contaminated area 15 km from Unit 4 were comparatively low. On the other hand, the effective doses around SNTS were below the public dose limit. These findings suggest that the environmental contamination and effective doses on the ground definitely decrease with decontamination such as removing surface soil, although the effective doses of the sampling points around CNPP in the present study were all over the public dose limit. Thus, the remediation of soil as a countermeasure could be an extremely effective method not only for areas around CNPP and SNTS but also for areas around the Fukushima Dai-ichi Nuclear Power Plant (FNPP), and external exposure levels will be certainly reduced. Long-term follow-up of environmental monitoring around CNPP, SNTS, and FNPP, as well as evaluation of the health effects in the population residing around these areas, could contribute to radiation safety and reduce unnecessary exposure to the public. PMID:23469013

  15. SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011

    Eibling, R.

    2011-09-28

    Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the

  16. Accumulation of transuranic elements in the aquatic biota of the Belarusian sector of contaminated area near the Chernobyl nuclear power plant - Accumulation of transuranic elements in aquatic biota of Belarusian sector of contaminated area of Chernobyl nuclear power plant

    Golubev, Alexander; Mironov, Vladislav [International Sakharov Environmental University. Box 220070, 23 Dolgobrodskaya Street, Minsk, 220070 (Belarus)

    2014-07-01

    The evolution of nuclear contamination of Belarus territory after Chernobyl accident includes the four stages: 1. Iodine-neptunium stage, caused mainly by short-lived radionuclides {sup 131}I, {sup 239}Np and others with a half-life period of several weeks; II. Intermediate stage, caused by radionuclides with a half-life period of a year ({sup 144}Ce, {sup 106}Ru, {sup 134}Cs, etc.); III. Strontium-cesium stage, caused by {sup 90}Sr and {sup 137}Cs with a half-life period of about 30 years; IV. Plutonium-americium, caused by long-lived α-emitting radionuclides {sup 241}Am (period of half-life of 432 years) and {sup 239+240}Pu, having high radio and chemo-toxicity. According to forecasts, activity of {sup 241}Am to 2050 year will increase by 2.5 times and it will be the most important dose-related factor for the aquatic biota within the Chernobyl accident zone. In 2002 - 2008 years we have studied the accumulation of trans-uranic elements (TUE, {sup 241}Am, {sup 239+240}Pu) in basic components of water body ecosystems within the Chernobyl zone - non-flowing Perstok Lake, weak-flowing Borschevka flooding and small Braginka River. Among investigated components are water, bottom sediments, submerged macrophytes (Ceratophyllum submersum, Hydrocharis morsus-ranae, Lemna minor, Nuphar lutea, Stratiotes aloides), emergent macrophytes (Typha spp.), shellfish and fish. In the soil cover in the vicinity of the Perstok Lake activity of {sup 241}Am at present is equivalent to 300 - 600 Bq.kg{sup -1}, that is the basic source of its income to the lake. Radionuclides mobility in the water environment is higher than in the soil, that facilitates the rapid incorporation of {sup 241}Am to the trophic nets of water bodies and its removal by near-water animals in the terrestrial biotopes, including outside Chernobyl zone. Thus, the activity of {sup 241}Am in bottom sediments in the Perstok Lake and Borschevka flooding in 2008 year reach respectively 324 and 131 Bq.kg{sup -1}, and the

  17. Saltstone Vault Classification Samples Modular Caustic Side Solvent Extraction Unit/Actinide Removal Process Waste Stream April 2011

    Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock and Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B and W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B and W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B and W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most

  18. Natural and anthropogenic radionuclides in Brazilian commercial dog food: preliminary results

    Cavalcante, Fernanda; Pecequilo, Brigitte R.S. [Instituto de Pesquisas Energeticas e Nucleares - IPEN, Av. Professor Lineu Prestes 2242, 05508-000 Sao Paulo (Brazil)

    2014-07-01

    spigot. These samples, after resting for 30 days to ensure secular equilibrium, were placed in an extended range high-purity germanium detector with 40% relative efficiency, and the acquired spectra are analyzed using the InterWinner 6.0 software. The results have shown no concentrations of artificial radionuclides, such as Cs-137, Co-60, Ru-106, Ru-103, I-131 and Am-241, whilst the concentrations of natural radionuclides varied from 1.17 ± 0.31 up to 5.01 ± 0.38 Bq/kg for Ra-226; from 1.20 ± 0.50 up to 8.07 ± 0.96 for Th-232 and from 212.90 ± 10.80 up to 377.00 ± 17.78 for K-40. Further, the study will be extended to a larger number of dog food brands and also to cat food brands available in Brazil and eventual radiological consequences of absorbed dose will be assessed. (authors)

  19. Natural and anthropogenic radionuclides in Brazilian commercial dog food: preliminary results

    spigot. These samples, after resting for 30 days to ensure secular equilibrium, were placed in an extended range high-purity germanium detector with 40% relative efficiency, and the acquired spectra are analyzed using the InterWinner 6.0 software. The results have shown no concentrations of artificial radionuclides, such as Cs-137, Co-60, Ru-106, Ru-103, I-131 and Am-241, whilst the concentrations of natural radionuclides varied from 1.17 ± 0.31 up to 5.01 ± 0.38 Bq/kg for Ra-226; from 1.20 ± 0.50 up to 8.07 ± 0.96 for Th-232 and from 212.90 ± 10.80 up to 377.00 ± 17.78 for K-40. Further, the study will be extended to a larger number of dog food brands and also to cat food brands available in Brazil and eventual radiological consequences of absorbed dose will be assessed. (authors)

  20. Accumulation of transuranic elements in the aquatic biota of the Belarusian sector of contaminated area near the Chernobyl nuclear power plant - Accumulation of transuranic elements in aquatic biota of Belarusian sector of contaminated area of Chernobyl nuclear power plant

    The evolution of nuclear contamination of Belarus territory after Chernobyl accident includes the four stages: 1. Iodine-neptunium stage, caused mainly by short-lived radionuclides 131I, 239Np and others with a half-life period of several weeks; II. Intermediate stage, caused by radionuclides with a half-life period of a year (144Ce, 106Ru, 134Cs, etc.); III. Strontium-cesium stage, caused by 90Sr and 137Cs with a half-life period of about 30 years; IV. Plutonium-americium, caused by long-lived α-emitting radionuclides 241Am (period of half-life of 432 years) and 239+240Pu, having high radio and chemo-toxicity. According to forecasts, activity of 241Am to 2050 year will increase by 2.5 times and it will be the most important dose-related factor for the aquatic biota within the Chernobyl accident zone. In 2002 - 2008 years we have studied the accumulation of trans-uranic elements (TUE, 241Am, 239+240Pu) in basic components of water body ecosystems within the Chernobyl zone - non-flowing Perstok Lake, weak-flowing Borschevka flooding and small Braginka River. Among investigated components are water, bottom sediments, submerged macrophytes (Ceratophyllum submersum, Hydrocharis morsus-ranae, Lemna minor, Nuphar lutea, Stratiotes aloides), emergent macrophytes (Typha spp.), shellfish and fish. In the soil cover in the vicinity of the Perstok Lake activity of 241Am at present is equivalent to 300 - 600 Bq.kg-1, that is the basic source of its income to the lake. Radionuclides mobility in the water environment is higher than in the soil, that facilitates the rapid incorporation of 241Am to the trophic nets of water bodies and its removal by near-water animals in the terrestrial biotopes, including outside Chernobyl zone. Thus, the activity of 241Am in bottom sediments in the Perstok Lake and Borschevka flooding in 2008 year reach respectively 324 and 131 Bq.kg-1, and the activity of 241Am in macrophytes of the Perstok Lake at the same year was 1,0 - 3,7 Bq.kg-1. In

  1. Patterns of Cs-137 and Sr-90 distribution in conjugated landscape systems

    Korobova, E.

    2012-04-01

    The main goal of the study was to reveal spatial patterns of 137Cs and 90Sr distribution in soils and plants of conjugated landscapes and to use 137Cs as a tracer for natural migration and accumulation processes in the environment. The studies were based on presumptions that: 1) the environment consisted of interrelated bio- and geochemical fields of hierarchical structure depending on the level and age of factors responsible for spatial distribution of chemical elements; 2)distribution of technogenic radionuclides in natural landscapes depended upon the location and type of the initial source and radionuclide involvement in natural pathways controlled by the state and mobility of the typomorphic elements and water migration. Case studies were undertaken in areas subjected to contamination after the Chernobyl accident and in the estuary zones of the Yenisey and Pechora rivers. First observations in the Chernobyl remote zone in 1987-1989 demonstrated relation between the dose rate, 137Cs, 134Cs, 144Ce, 106Ru, 125Sb in soil cover and the location of the measured plot in landscape toposequence. Later study of 137Cs and 90Sr concentration and speciation confirmed different patterns of their distribution dependent upon the radioisotope, soil features and vegetation cover corresponding to the local landscape and landuse structure. Certain patterns in distribution and migration of 137Cs and 90Sr in soils and local food chain were followed in private farms situated in different landscape position [1]. Detailed study of 137Cs activity in forested site with a pronounced relief 20 and 25 years after the Chernobyl accident showed its stable polycentric structure in soils, mosses and litter which was sensitive to meso- and micro-relief features [2]. Radionuclide contamination of the lower Yenisey and Pechora studied along meridian landscape transects proved both areas be subjected to global 137Cs pollution while the Yenisey floodplain received additional regional contamination

  2. Establishment of methodology for determination of {sup 93}Zr in radioactive wastes by Liquid Scintillation Counting (LSC) and Inductively Coupled Plasma Mass Spectrometry (ICP-MS); Estabelecimento de metodologia para determinacao de {sup 93}Zr em rejeitos radioativos por Espectrometria de Cintilacao Liquida (LSC) e Espectrometria de Massa com Plasma Indutivamente Acoplado (ICP-MS)

    Oliveira, Thiago Cesar de

    2014-06-01

    The zirconium-93 is a long-lived pure β-particle-emitting radionuclide produced from {sup 235}U fission and from neutron activation of the stable isotope {sup 92}Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, {sup 93}Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of {sup 93}Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. The aim of this work was to develop a selective radiochemical separation methodology for the determination of {sup 93}Zr in nuclear waste and analyze it by Liquid Scintillation Counting (LSC) and Inductively Coupled Plasma Mass Spectrometry (ICP-MS). To set up the radiochemical separation procedure for zirconium, a tracer solution of {sup 95}Zr and its 724 keV γ-ray measurements by γ- spectrometry were used in order to follow the behavior of zirconium during the radiochemical separation. For the LSC technique a {sup 55}Fe solution, which is one of the major interfering measures zirconium, was used to verify the decontamination factor during the separation process. The efficiency detection for {sup 63}Ni was used to determination of {sup 93}Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq 1{sup −1} was obtained for {sup 63}Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for Optiphase Hisafe 3 cocktail. For the ICP-MS technique a zirconium stable solution was used to verify the zirconium behavior and recovery during radiochemical separation and a solution of Ba, Co, Eu, Fe, Mn, Nb, Sr and Y was used to verify the decontamination factor during the separation process. A standard solution {sup 93}Nb as isotope for determining the {sup 93}Zr by ICP-MS was used for calibration and analysis. The detection limit of 0.039 ppb was obtained for the standard

  3. Formation of Microbial Mats and Salt in Radioactive Paddy Soils in Fukushima, Japan

    Kazue Tazaki

    2015-12-01

    Full Text Available Coastal areas in Minami-soma City, Fukushima, Japan, were seriously damaged by radioactive contamination from the Fukushima Daiichi Nuclear Power Plant (FDNPP accident that caused multiple pollution by tsunami and radionuclide exposure, after the Great East Japan Earthquake, on 11 March 2011. Some areas will remain no-go zones because radiation levels remain high. In Minami-soma, only 26 percent of decontamination work had been finished by the end of July in 2015. Here, we report the characterization of microbial mats and salt found on flooded paddy fields at Karasuzaki, Minami-soma City, Fukushima Prefecture, Japan which have been heavily contaminated by radionuclides, especially by Cs (134Cs, 137Cs, 40K, Sr (89Sr, 90Sr, and 91 or 95Zr even though it is more than 30 km north of the FDNPP. We document the mineralogy, the chemistry, and the micro-morphology, using a combination of micro techniques. The microbial mats were found to consist of diatoms with mineralized halite and gypsum by using X-ray diffraction (XRD. Particular elements concentrated in microbial mats were detected using scanning electron microscopy equipped with energy dispersive spectroscopy (SEM-EDS and X-ray fluorescence (XRF. The objective of this contribution is to illustrate the ability of various diatoms associated with minerals and microorganisms which are capable of absorbing both radionuclides and stable isotopes from polluted paddy soils in extreme conditions. Ge semiconductor analysis of the microbial mats detected 134Cs, 137Cs, and 40K without 131I in 2012 and in 2013. Quantitative analysis associated with the elemental content maps by SEM-EDS indicated the possibility of absorption of radionuclide and stable isotope elements from polluted paddy soils in Fukushima Prefecture. In addition, radionuclides were detected in solar salts made of contaminated sea water collected from the Karasuzaki ocean bath, Minami-soma, Fukushima in 2015, showing high Zr content associated

  4. I. Nuclear Production Reaction and Chemical Isolation Procedure for Americium-240 II. New Superheavy Element Isotopes: Plutonium-242(Calcium-48,5n)(285)114

    Ellison, Paul Andrew

    2011-12-01

    Part I discusses the study of a new nuclear reaction and chemical separation procedure for the production of 240Am. Thin 242Pu, natTi, and natNi targets were coincidently activated with protons from the 88-Inch Cyclotron, producing 240Am, 48V, and 57Ni, respectively. The radioactive decay of these isotopes was monitored using high-purity Ge gamma ray detectors in the weeks following irradiation. The excitation function for the 242 Pu(p, 3n)240Am nuclear reaction was measured to be lower than theoretical predictions, but high enough to be the most viable nuclear reaction for the large-scale production of 240 Am. Details of the development of a chemical separation procedure for isolating 240Am from proton-irradiated 242Pu are discussed. The separation procedure, which includes two anion exchange columns and two extraction chromatography columns, was experimentally investi- gated using tracer-level 241Am, 239Pu, and model proton-induced fission products 95Zr, 95Nb, 125Sb, and 152Eu. The separation procedure was shown to have an Am/Pu separation factor of >2x10 7 and an Am yield of ˜70%. The separation procedure was found to purify the Am sample from >99.9% of Eu, Zr, Nb, and Sb. The procedure is well suited for the processing of ˜1 gram of proton-irradiated 242Pu to produce a neutron-induced fission target consisting of tens of nanograms of 240Am. Part II describes the use of the Berkeley Gas-filled Separator at the Lawrence Berkeley National Laboratory 88-Inch Cyclotron for the study of the 242Pu(48Ca,5n)285114 nuclear re- action. The new, neutron-deficient, superheavy element isotope 285114 was produced in 48Ca irradiations of 242Pu targets at a center-of-target beam energy of 256 MeV ( E* = 50 MeV). The alpha decay of 285114 was followed by the sequential alpha decay of four daughter nuclides, 281Cn, 277Ds, 273Hs, and 269 Sg. 265Rf was observed to decay by spontaneous fission. The measured alpha-decay Q-values were compared with those from a macroscopic

  5. Design of the Capsule (13M-01K) for Irradiation of Fuel Cladding Materials in HANARO

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear-grade zirconium alloys contain more than 95% Zr, and therefore most of their properties are similar to those of pure zirconium. ZIRLO material, used in fuel rod cladding, structural and flow mixing grids, instrumentation tubes, and guide thimbles, increases margin-to-fuel-rod-corrosion limits and enhances fuel assembly structural stability. The demonstrated corrosion resistance and enhanced structural stability of ZIRLO cladding enable longer cycle lengths at higher temperatures without reducing operating margins. An instrumented capsule (13M-01K) was designed and fabricated for evaluation of the neutron irradiation properties of Zirlo material, which is commonly used for cladding of nuclear fuel. This capsule is now being irradiated for 2 cycles at CT test hole of HANARO, which was started at Jan 27 and will be ended at Mar 31, 2014. The structure of the capsule was based on the previous capsule (11M-22K capsule) which was successfully irradiated at the same hole of HANARO. In the capsule, 182 specimens such as tensile specimens of plate type and ring type specimens were placed. Most of them are made of Zirlo, but a few are HANA material that is developed in KAERI. The irradiation test was requested by 4 universities including Dong-Kook and Han-Yang etc. The capsule is composed of 5 layers, each of which had Al holder containing several specimens and an independent electric heater, thermocouples etc. During the irradiation test, temperatures of the specimens and fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule is irradiated for 2 cycles (28 days) at the CT test hole of HANARO of a 30MW thermal output at 350-390 .deg. C up to a fast neutron fluence of 9.6Χ10{sup 20} (n/cm{sup 2}) (E>1.0 Mev). In this capsule, the two kinds of irradiation tests are performed at quite different temperatures. At upper 3 layers of the capsule, specimens for irradiation at high

  6. Separation of {sup 90}Nb from zirconium target for application in immuno-PET

    Radchenko, V.; Roesch, F. [Mainz Univ. (Germany). Inst. of Nuclear Chemistry; Filosofov, D.V.; Bochko, O.K.; Lebedev, N.A.; Rakhimov, A.V. [Joint Institute for Nuclear Research, Dubna, Moscow Region (Russian Federation). Dzhelepov Laboratory of Nuclear Problems; Hauser, H.; Eisenhut, M. [German Cancer Research Center, Heidelberg (Germany). Radiopharmaceutical Chemistry; Aksenov, N.V.; Bozhikov, G.A. [Joint Institute for Nuclear Research, Dubna, Moscow Region (Russian Federation). Flerov Laboratory of Nuclear Reactions; Ponsard, B. [Belgian Nuclear Research Centre (SCKCEN), Mol (Belgium). Radioisotopes and NTD Silicon Production

    2014-07-01

    Fast progressing immuno-PET asks to explore new radionuclides. One of the promising candidates is {sup 90}Nb. It has a half-life of 14.6 h that allows visualizing and quantifying biological processes with medium and slow kinetics, such as tumor accumulation of antibodies and antibodies fragments or drug delivery systems and nanoparticles. {sup 90}Nb exhibits a positron branching of 53% and an average kinetic energy of emitted positrons of E{sub mean} = 0.35 MeV. Currently, radionuclide production routes and NbV labeling techniques are explored to turn this radionuclide into a useful imaging probe. However, efficient separation of {sup 90}Nb from irradiated targets remains in challenge. Ion exchange based separation of {sup 90}Nb from zirconium targets was investigated in systems AG 1 x 8 - HCl/H{sub 2}O{sub 2} and UTEVA-HCl. {sup 95}Nb (t{sub 1/2} = 35.0 d), {sup 95}Zr (t{sub 1/2} = 64.0 d) and {sup 92m}Nb (t{sub 1/2} = 10.15 d) were chosen for studies on distribution coefficients. Separation after AG 1 x 8 anion exchange yields 99% of {sup 90/95}Nb. Subsequent use of a solid-phase extraction step on UTEVA resin further decontaminates {sup 90/95}Nb from traces of zirconium with yields 95% of {sup 90/95}Nb. A semi-automated separation takes one hour to obtain an overall recovery of {sup 90/95}Nb of 90%. The amount of Zr was reduced by factor of 10{sup 8}. The selected separation provides rapid preparation (< 1 h) of high purity {sup 90}Nb appropriate for the synthesis of {sup 90}Nb-radiopharmaceuticals, relevant for purposes of immuno-PET. Applying the radioniobium obtained, {sup 90/95}Nb-labeling of a monoclonal antibody (rituximab) modified with desferrioxamine achieved labeling yields of > 90% after 1 h incubation at room temperature. (orig.)

  7. Human biokinetic data and a new compartmental model of zirconium - A tracer study with enriched stable isotopes

    Greiter, Matthias B., E-mail: matthias.greiter@helmholtz-muenchen.de; Giussani, Augusto, E-mail: AGiussani@BfS.de; Hoellriegl, Vera, E-mail: vera.hoellriegl@helmholtz-muenchen.de; Li Weibo, E-mail: wli@helmholtz-muenchen.de; Oeh, Uwe, E-mail: uwe.oeh@helmholtz-muenchen.de

    2011-09-01

    Biokinetic models describing the uptake, distribution and excretion of trace elements are an essential tool in nutrition, toxicology, or internal dosimetry of radionuclides. Zirconium, especially its radioisotope {sup 95}Zr, is relevant to radiation protection due to its production in uranium fission and neutron activation of nuclear fuel cladding material. We present a comprehensive set of human data from a tracer study with stable isotopes of zirconium. The data are used to refine a biokinetic model of zirconium. Six female and seven male healthy adult volunteers participated in the study. It includes 16 complete double tracer investigations with oral ingestion and intravenous injection, and seven supplemental investigations. Tracer concentrations were measured in blood plasma and urine collected up to 100 d after tracer administration. The four data sets (two chemical tracer forms in plasma and urine) each encompass 105-240 measured concentration values above detection limits. Total fractional absorption of ingested zirconium was found to be 0.001 for zirconium in citrate-buffered drinking solution and 0.007 for zirconium oxalate solution. Biokinetic models were developed based on the linear first-order kinetic compartmental model approach used by the International Commission on Radiological Protection (ICRP). The main differences of the optimized systemic model of zirconium to the current ICRP model are (1) recycling into the transfer compartment made necessary by the observed tracer clearance from plasma, (2) different parameters related to fractional absorption for each form of the ingested tracer, and (3) a physiologically based excretion pathway to urine. The study considerably expands the knowledge on the biokinetics of zirconium, which was until now dominated by data from animal studies. The proposed systemic model improves the existing ICRP model, yet is based on the same principles and fits well into the ICRP radiation protection approach. - Research

  8. Fission Product Yields from Fission Spectrum n+239Pu for ENDF/B-VII.1

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small - especially for 99Mo - we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the

  9. Discovery of Self-Sustained 235-U Fission Causing Sunlight by Padmanabha Rao Effect

    Rao, M. A. Padmanabha

    2013-07-01

    For the first time in solar physics, this paper reports a comprehensive study how 235Uranium fission causes Sunlight by the atomic phenomenon, Padmanabha Rao Effect against the theory of fusion. The first major breakthrough lies in identifying as many as 153 solar lines in the Bharat Radiation range from 12.87 to 31 nm reported by various researchers since 1960s. The Sunlight phenomenon is explained as follows. For example, the energy equivalence 72.48 eV of the most intense 17.107 nm emission in the middle of solar spectrum is the energy lost by β, γ, or X-ray energy of a fission product while passing through core-Coulomb space. This energy loss is the Bharat Radiation energy that cause EUV, UV, visible, and near infrared emissions on valence excitation. From vast data of emissions and energies of various fission products, 606.31 keV β (Eβmax) energy of 131I was chosen as the source of 17.107 nm emission. For the first time a typical Bharat Radiation spectrum was observed when plotted energy loss against β, γ, or X-ray energies of fission products supposedly present in solar flare and atmosphere : 113Xe, 131I, 137Cs, 95Zr, 144Cs, 134I, 140Ba, 133I, 140La, 133In etc that caused solar lines. Consistent presence of a sharp line for four months in AIA spectral EUV band at 335A exemplifies self-sustained uranium fission from a small site appeared in SDO/AIA image at 304A. Sun's dark spot is explained as a large crater formed on Sun's core surface as a result of fission reaction that does not show any emission since fission products would be thrown away from the site during fission. Purely the same Sun's core material left over at the site after fission reaction devoid of fission products and any emission seems to be the familiar dark Matter. This could be the first report on the existence of Sun's Dark Matter.

  10. Establishment of methodology for determination of 93Zr in radioactive wastes by Liquid Scintillation Counting (LSC) and Inductively Coupled Plasma Mass Spectrometry (ICP-MS)

    The zirconium-93 is a long-lived pure β-particle-emitting radionuclide produced from 235U fission and from neutron activation of the stable isotope 92Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, 93Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of 93Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. The aim of this work was to develop a selective radiochemical separation methodology for the determination of 93Zr in nuclear waste and analyze it by Liquid Scintillation Counting (LSC) and Inductively Coupled Plasma Mass Spectrometry (ICP-MS). To set up the radiochemical separation procedure for zirconium, a tracer solution of 95Zr and its 724 keV γ-ray measurements by γ- spectrometry were used in order to follow the behavior of zirconium during the radiochemical separation. For the LSC technique a 55Fe solution, which is one of the major interfering measures zirconium, was used to verify the decontamination factor during the separation process. The efficiency detection for 63Ni was used to determination of 93Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq 1−1 was obtained for 63Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for Optiphase Hisafe 3 cocktail. For the ICP-MS technique a zirconium stable solution was used to verify the zirconium behavior and recovery during radiochemical separation and a solution of Ba, Co, Eu, Fe, Mn, Nb, Sr and Y was used to verify the decontamination factor during the separation process. A standard solution 93Nb as isotope for determining the 93Zr by ICP-MS was used for calibration and analysis. The detection limit of 0.039 ppb was obtained for the standard solution of zirconium. Then, the protocol was applied to low level waste (LLW

  11. Leaching study of nuclear melt glass: Part I

    Ground samples of three nuclear melt glasses from underground nuclear explosions at the Nevada Test Site (NTS) were leached at 250C with natural ground water from NTS. Using our dynamic single-pass flow-through leaching system we monitored the release of radionuclides from the glasses during 420 days of leaching. We continually flowed the ground water over the melt glass at flow rates of 185 ml/day for half of the samples and 34 ml/day for the rest. Leachate solutions were collected continuously, and composite samples, collected on days 1, 2, 3, 6, 11, 32, 38, 70, 120, 230 and 420, were analyzed using low-background Ge(Li) gamma spectrometers. For most of the radionuclides the leach rate decreased smoothly throughout the experiment. Except for 95Zr, 144Ce, and 155Eu, there was no difference between the fast (185 ml/day) and slow (34 ml/day) flow-rate leach rates. The measurable leach rates ranged from a high of 1 x 10-2 g-glass/m2 day for 22Na (slow flow-rate, day 1 in glass No. 2) to a low of 1 x 6-6 g-glass/m2 day for 54Mn (slow flow-rate, day 420 in glass No. 2). Most of the leach-rate values were about 5 x 10-4 g-glass/m2 day initially, decreasing to 5 x 10-6 g-glass/m2 day after 420 days of leaching. The leach rates had not leveled off by the end of the experiment and were, in general, continuing to decrease. From the activities in the leachate solutions, we determined the percent of the pre-leach activity leached from the melt glasses during the experiment. Only 124Sb, in one fast flow-rate channel of glass No. 2, exceeded 3% of the initial activity leached. The majority of the samples released less than 1% of the pre-leach activity for a given radionuclide. The percent activity released from the samples leached at the fast and slow flow rates were nearly equal

  12. Physical and chemical characteristics of radionuclide carriers the environment

    In spite of the nuclear energy advantages related to the smallest environmental impact due to protection of resources and reduction of greenhouse gas emissions, the environment has been affected by nuclear activities. In order to evaluate trends, to study transport processes and to predict consequences of radionuclide releases/leakage from a source and atmospheric fallout, information on the present levels of radioactive contamination, releases and source terms is essential. At present it is generally recognized that a realistic estimation of the long-term consequences of radioactive contamination is possible through the better understanding of physical and chemical processes of radionuclide migration in the environment. To assess the transfer and fate of radioactive contaminants in the atmosphere, terrestrial and marine environment the information on physical and chemical parameters of radionuclide carriers is required. The release of radionuclides associated with particles of different sizes and mineralogical composition into the environment can considerably affect their transport and bioavailability. Thus, after nuclear explosions different spreading velocities and downward movement of 137Cs, 95Zr+95Nb, 22Na aerosols and gaseous 14C(14CO2) were observed. In addition, low chemical reactivity is characteristic of highly insoluble, refractory oxides of uranium and plutonium formed in nuclear explosions, and they are very kinetically stable and remain in a form in which they are injected into environment for a long time. They were distinguished for their behaviour in the environment from radionuclides released into environment processing plants and laboratory research The emission of radiocesium from combustion of contaminated firewood can also contribute to the radiational situation in Lithuania. The analysis of activity concentrations, meteorological situation, types of particle are important for understanding the sources and possible impact on given location. Our

  13. ZZ KAFAX-F31, 150 and 12 Groups Cross Section Library in MATXS Format based on JEFF-3.1 for Fast Reactors

    1 - Description: Format: MATXS, 142 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 142 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-nat, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JEFF-3.1. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-F31 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JEFF-3.1. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 142 nuclide data (Table 1) processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-F31 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost

  14. ZZ KAFAX-E70, 150 and 12 Groups Cross Section Library in MATXS Format based on ENDF/B-VII.0 for Fast Reactors

    1 - Description: Format: MATXS, 144 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 144 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-113, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: ENDF/B-VII.0. Weighting spectrum: 300, 600, 900, 1200 k. The ZZ-KAFAX-E70 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on ENDF/B-VII.0. This library was originally generated for the KALIMER (Korea Advanced Liquid Metal Reactor) core analyses. It includes 144 nuclide data processed with the NJOY99.245 code patched with NEA020. The library can be used to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-E70 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy

  15. ZZ KAFAX-J33, 150 and 12 Groups Cross Section Library in MATXS Format based on JENDL-3.3 for Fast Reactors

    1 - Description: Format: MATXS, 136 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 136 Nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Pb-206, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JENDL-3.3. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-J33 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JENDL-3.3. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 136 nuclide data processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-J33 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost all the energy ranges, except between 1 and 10 keV in

  16. Modulation-Doped SrTiO3/SrTi1-xZrxO3 Heterostructures

    Kajdos, Adam Paul

    surface reconstruction from (1x1) to (2x1) to c(4x4) is correlated with a change from mixed SrO/TiO2 to pure TiO2 surface termination. It is argued that optimal cation stoichiometry is achieved for growth conditions within the XRD-defined growth window that result in a c(4x4) surface lattice. The development of a doped perovskite oxide semiconductor with a suitable conduction band offset is then discussed as the next necessary step towards realizing modulation-doped heterostructures. The SrTixZr1-x O3 solid solution is investigated for this purpose, with a focus on optimizing cation stoichiometry to allow for controlled doping. In particular, the hybrid MBE growth of SrTixZr1-xO3 thin films is explored using a metal-organic precursor for Zr, zirconium tert-butoxide (ZTB). The successful generation of 2DEGs by modulation doping of SrTiO3 is then demonstrated in SrTiO3/La:SrTi0.95Zr0.05O 3 heterostructures, and the electronic structure is studied by Shubnikov-de Haas analysis using multiple-subband models.

  17. Fission Product Yields from Fission Spectrum n+ 239Pu for ENDF/B-VII.1

    Chadwick, M. B.; Kawano, T.; Barr, D. W.; Mac Innes, M. R.; Kahler, A. C.; Graves, T.; Selby, H.; Burns, C. J.; Inkret, W. C.; Keksis, A. L.; Lestone, J. P.; Sierk, A. J.; Talou, P.

    2010-12-01

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for 99Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on

  18. Application of LiF for determining the gamma-radiation characteristics of the shut-down reactor

    Full text: The power of 60Co ∼1.25 MeV gamma-radiation source at the INP AS RUz is limited by 8 Gy/s, which does not satisfy several tasks of material science now. Therefore, we were first to suggest the irradiation of materials with gamma-rays of 0.1-7 MeV, which are emitted by the uranium fission products (41Ar, 135Xe, 125Xe, 125I,137Cs, 134Cs, 144Ce, 95Zr, 140Ba, 140La, 99Mo, 60Co) and l6N, 24Na, 28Al radio-nuclides in water during prophylactic shut-downs of our nuclear reactor WWR-SM. The gamma-dose rate kinetics was monitored with the ion current in ionization chambers KNK-53M fixed outside the reactor core from the stop-moment. The current kinetics comprised 4 steps with a high reproducibility at 2 and 0.5 μA, then 50 and 10 nA, each lasting for 1,10, 40 and up to 200 hours, according to the isotope life-times. LiF crystal is known as a thermal luminescence dosimeter of mixed radiations up to 100 Gy. Yet in this work the absorbed gamma-energy dose Dγ was determined by accumulation of the known stable structure defects in thin cleaved LiF crystals: by induced optical absorption and luminescence of F- and M-centers. The samples were irradiated in Al-containers filled with water to keep the temperature of ∼40 deg. C in the time range from 30 minutes to 150 hours. Optical absorption spectra were registered at spectrometer Specord M-40. Then the induced color center concentration was calculated by the Smakula relation, which is proportional to the absorbed dose Dγ. For a better reliability the photoluminescence center content was also determined. Selecting comparable close intensities of the induced absorption and luminescence bands obtained after irradiations of LiF references in the certified 60Co gamma-sources of the known gamma fluxes 0.7 and 7.5 Gy/s, the gamma-radiation intensity of the shut-down reactor was estimated in correlation with the ion current as 10 nA = 15 Gy/s. At short times of irradiation the linear dose dependence occurred for the